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	description	This invention claims priority from Provisional Application Ser. No. 60/889,312, filed Feb. 12, 2007. This invention was made with government support under Contract No. DE-FC07-051D14636 awarded by the Department of Energy. The government has certain rights in this invention. 1. Field of the Invention This invention relates to water-cooled nuclear reactors and more particularly to apparatus for improving the distribution of coolant entering the core of water-cooled nuclear reactors. 2. Description of the Prior Art The primary side of nuclear reactor power generating systems which are cooled with water under pressure comprises a closed circuit which is isolated from and in heat exchange relationship with a secondary side for the production of useful energy. The primary side comprises the reactor vessel enclosing a core internals structure that supports a plurality of fuel assemblies containing fissile material, the primary circuit within heat exchange steam generators, the inner volume of a pressurizer, pumps and pipes for circulating pressurized water; the pipes connecting each of the steam generators and pumps to the reactor vessel independently. Each of the parts of the primary side comprising a steam generator, a pump and a system of pipes which are connected to the vessel form a loop of the primary side. The primary side is also connected to auxiliary circuits, including a circuit for volumetric and chemical monitoring of the pressurized water. The auxiliary circuit, which is arranged branching from the primary circuit, makes it possible to maintain the quantity of water in the primary circuit by replenishing, when required, with measured quantities of water, and to monitor the chemical properties of the coolant water, particularly its content of boric acid, which is important to the operation of the reactor. The average temperature of the core components during full power reactor operation is approximately 580° F. (304° C.). Periodically, it is necessary to shut down the reactor system for maintenance and to gain access to the interior side of the pressure vessel. During such an outage, the internal components of the pressure vessel can cool to a temperature of approximately 50° F. (10° C.). The internal components of a pressure vessel typically consist of upper and lower internals. The upper internals include a control rod guide tube assembly, support columns, conduits for instrumentation which enter the reactor through the closure head, and a fuel assembly alignment structure, referred to as the upper core plate. The lower internals include a core support structure referred to as the core barrel, a core shroud that sits inside the core barrel and converts the circular interior of the barrel to a stepped pattern that substantially corresponds to the perimeter profile of the fuel assemblies that constitute the core supported between the lower core support plate and the upper core support plate. Generally, the reactor vessel is cylindrical having a hemispherical lower end. The core barrel is connected to the interior walls of the reactor vessel at or adjacent to the area where the cylindrical and hemispherical portions of the reactor vessel meet. Below the main core support, i.e. the core barrel that is capped at its lower end with the lower core support, the hemispherical vessel defines a lower head or lower plenum. A generally annular downcomer surrounds the reactor core barrel between the core barrel and an inner wall of the reactor vessel. Cooling fluid, typically water, is pumped into this annular downcomer. The coolant fluid circulates downward into the lower plenum. The hemispherical shape of the lower plenum assists in evenly circulating the coolant fluid therein. A plurality of reactor core coolant inlet openings are located on the underside of the lower core support plate. Coolant flows from the lower plenum, into the core coolant inlet openings and upwardly into the core to cool the fuel assemblies. In order to maintain adequate and uniform cooling throughout the core, it is important that a uniform coolant flow and pressure be maintained across all of the reactor core coolant inlet openings in the lower core support plate. Non-uniform coolant pressure or flow causes uneven coolant flow into the core, which results in uneven cooling of the fuel assemblies of the core. Uneven fuel assembly cooling may force the entire core to be derated to accommodate “hot assembly” locations. Non-uniform coolant flow and pressure may result in vortices or other flow disruptions to form in the coolant fluid circulating in the lower plenum. It is desirable to provide core monitoring instrumentation within the core of a nuclear reactor. Traditionally, the leads connecting such instrumentation to the exterior of the reactor exit the reactor vessel through a central portion of the lower hemispherical portion of the reactor vessel. A plurality of conduits extending from the underside of the lower core plate to the interior walls of the lower hemispherical portion of the reactor vessel carry the instrumentation lines through the lower plenum. The presence of the conduits in the lower plenum assists in maintaining even coolant flow within the lower plenum and disrupting the formation of vortices in the circulating coolant fluid. Such vortices disrupt coolant flow and produce low pressure areas at the core coolant inlets which they intersect. In newer reactors, it has become desirable for any instrumentation conduits to exit the reactor vessel other than through the lower plenum. It has been found that the absence of instrumentation conduits from the lower plenum permits vortices to form in the circulating coolant in the lower plenum. U.S. Pat. No. 5,267,285 issued Nov. 30, 1993 and assigned to the assignee of this invention, suggested the use of one or more spaced parallel plates, supported in the lower plenum parallel to the lower support plate with holes for the passage of coolant, as a means to suppress vortices. With the advent of larger passive plants with larger cores it became evident that further means were necessary to improve the distribution of coolant flow in the lower plenum to assure uniform coolant flow and pressure were maintained across all of the reactor core coolant inlet openings in the lower core support plate. Accordingly, there is a further need to improve the design of the lower reactor vessel plenum to assure this uniform flow and pressure is maintained. These and other objects are achieved by employing a cylindrical reactor pressure vessel having a lower head and a lower core support plate. A cylinder having a vertical wall with a plurality of holes extending therethrough and an upper edge is supported from a plurality of locations around the lower head with the upper edge of the cylinder proximate the lower core support plate so that the majority of coolant flow entering the reactor pressure vessel and down the annulus between the cylindrical reactor pressure vessel walls and the core barrel passes through the holes in the vertical wall of the cylinder on route to the core inlet holes in the lower core support plate which is supported by the lower end of the core barrel. Desirably, the upper edge of the cylinder is spaced below a bottom surface of the lower core support plate. Preferably a circumferential rib extends radially inward from the interior of the vertical wall of the cylinder to stiffen the wall so it can withstand the pressure of the flowing coolant. In one embodiment the circumferential rib is formed slightly above a center of the height of the vertical wall of the cylinder and the holes in the vertical wall are formed in a first and second pattern. Preferably, the holes in the vertical wall of the cylinder are substantially the same size, and the first pattern of holes is above the rib and the second pattern of holes is below the rib. Desirably the first pattern is substantially, circumferentially continuous and the second pattern is not substantially, circumferentially continuous. In one embodiment a plurality of pairs of spaced vertical slots, extending from a bottom of the cylinder vertically upward, are cut in a lower portion of the vertical wall of the cylinder, with each pair of vertical slots forming an attachment leg that is connected to the lower head. Preferably the circumferential distance between some of the pairs of spaced vertical slots differ from the distance between others of the pairs of spaced vertical slots and desirably the slots are symmetric across a diameter of the cylinder. Preferably the connection locations of the attachment legs to the lower plenum are not vertically aligned with the attachment locations of the core barrel to the reactor pressure vessel. In one embodiment six to eight attachment legs are provided and the slots extend up vertically to just below the elevation of the rib. Desirably the non-continuous pattern of holes of the second pattern are separated by the attachment legs. In another embodiment the holes in the first pattern and the holes in the second pattern are substantially aligned in circumferential rows and the number of rows of the second pattern are larger than the number of rows of the first pattern. Preferably, each row of the first pattern is nested in another row of the first pattern and each row of the second pattern is nested in another row of the second pattern. Preferably, the cylinder substantially closes off the space between the annulus and a bottom portion of the lower head and the holes in the vertical wall of the cylinder are arranged so that a first portion of the coolant is directed directly up through the plurality of holes in the lower core support plate and a second portion of the coolant is directed downward toward the bottom portion of the lower head and up through the vortice suppression plates. Referring now to the drawings, FIG. 1 shows a simplified nuclear reactor primary system, including a generally cylindrical reactor pressure vessel (10) having a closure head (12) enclosing a nuclear core (14). A liquid reactor coolant, such as water, is pumped into the vessel (10) by pump (16) through the core (14) where heat energy is absorbed and is discharged to a heat exchanger (18), typically referred to as a steam generator, in which heat is transferred to a utilization circuit (not shown), such as a steam driven turbine generator. The reactor coolant is then returned to the pump (16) completing the primary loop. Typically, a plurality of the above-described loops are connected to a single reactor vessel (10) by reactor coolant piping (20). An exemplary reactor design is shown in more detail in FIG. 2. In addition to a core (14) comprised of a plurality of parallel, vertical co-extending fuel assemblies (22), for purposes of this description, the other vessel internal structures can be divided into the lower internals (24) and the upper internals (26). In conventional designs, the lower internals function is to support, align and guide core components and instrumentation, as well as direct flow within the vessel. The upper internals restrain or provide a secondary restraint for the fuel assemblies (22), and support and guide instrumentation and components, such as control rods (28). In the exemplary reactor shown in FIG. 2, coolant enters the vessel (10) through one or more inlet nozzles (30), flows downward through an annulus between the vessel (10) and the core barrel (32), is turned 180° in a lower plenum (34), passes upwardly through a lower core support plate (36) upon which the fuel assemblies (22) are seated, and through and about the assemblies. The coolant flow through the core and surrounding area is typically large, on the order of 400,000 gallons per minute at a velocity of approximately 20 feet per second. The resulting pressure drop and frictional forces tend to cause the fuel assemblies to rise, which movement is restrained by the upper internals (26), including a circular upper core plate (38). Coolant exiting the core (14) flows along the underside of the upper core plate (38) and upwardly through a plurality of perforations. The coolant then flows upwardly and radially to one or more outlet nozzles (40). Rectilinearly moveable control rods (28) typically include a drive shaft and a spider assembly of neutron poison rods that are guided through the upper internals (26) and into aligned fuel assemblies (22) by control rod guide tubes (48). The guide tubes (48) are fixedly joined to the upper support assembly (44) and connected by a split pin force fit into the top of the upper core plate (38). The pin configuration provides for ease of guide tube assembly or replacement if ever necessary and assures that the core loads, particularly under seismic or other high loading accident conditions, are taken primarily by the support columns (46) and not the guide tubes (48). This assists in retarding guide tube deformation under accident conditions which could detrimentally affect control rod insertion capability. In larger plants it is desirable to further refine the flow pattern in the lower plenum (34) to assure that a constant flow rate and pressure is maintained over the entire surface of the lower core support plate (36). To accomplish that objective this invention employs a flow skirt schematically shown in FIG. 2 by reference character (50) and shown in more detail in FIGS. 3-6. The flow skirt (50) is a perforated cylinder structure in the lower reactor vessel plenum (34) that channels the coolant exiting the annulus between the vessel (10) and the core barrel (32) through flow holes in the flow skirt (50) and is attached to the reactor vessel bottom head (52) by welding an integral attachment leg (66) to a land (56) on the reactor vessel bottom head (52) at circumferentially spaced locations around the bottom head. The mounting locations of the flow skirt (50) on the lands (56), circumferentially spaced around the bottom head (52) are diametrically symmetrical, though not evenly spaced, so that the mounting locations do not align vertically with the attachment positions of the core barrel to the interior of the reactor vessel wall. FIG. 3 illustrates the connection of the flow skirt (50) to the lower vessel head (52) and the relative spacing of the flow skirt (50) with respect to the lower core support plate (36). A space (70) is maintained between a flow skirt (50) and the lower core support plate (36) to provide for relative movement of the core barrel, e.g., in the event of a seismic event, and differential thermal expansion, so the flow skirt is not damaged. The holes (72) in the flow skirt (50) are all approximately the same size and are arranged in two separate patterns. The first pattern (62) extends between an upper edge (68) of the flow skirt and a horizontal rib that forms a circumferential recess (58) in the outer surface of the flow skirt (50). The holes (72) in the first hole pattern (62) are formed in two circumferentially continuous rows that are nested together. The second hole pattern (64) is formed from four circumferentially discontinuous rows of holes (72) that are nested together on the surface of the vertical wall of the flow skirt (50) below the recess (58). The circumferential extent of the rows in the hole pattern (64) is interrupted by the attachment legs (66), which are shown in FIGS. 4, 5 and 6. A circumferential lip (60) extends around the flow skirt (60) and defines an increase in the thickness of the vertical wall of the flow skirt (50) below the lip (60) that strengthens the flow skirt and supports it against the forces of coolant flow. FIG. 4 shows an elevational view of a quarter of the lower plenum (34) and provides a view of one of the radial keys that support the core barrel on the vessel (10) and its positioning relative to the support leg (66). The radial keys (74) are supported on the cardinal axis of the vessel (10) while the support legs are designed to be offset from the cardinal access. In all other respects, the flow skirt is shown in FIG. 4 as previously described with respect to FIG. 3. The lower core support plate (36) has an access port (76) whose cover can be removed to inspect the internals below the lower core support plate during plant outages. Typically, there are six to eight attachment legs (66) supporting the flow skirt (50) on the lower head (52). FIG. 5 shows a flat pattern view of the outside surface of the vertical wall of the flow skirt (50). FIG. 5 provides a good view of the substantially circumferentially continuous pattern (62) having two rows of holes (72) nested together. FIG. 5 also provides a good view of the substantially circumferentially discontinuous hole pattern (64), which comprises four rows of the holes (72) nested together and circumferentially interrupted by the attachment legs (66). FIG. 6 provides an isometric view of the portion of the interior of the flow skirt (50) showing the circumferential stiffening rib (58) which extends radially inward to strengthen the flow skirt (50). Like reference characters have been used among the several figures to identify corresponding components. Referring back to FIGS. 2 and 3, the coolant enters through the inlet (30) into the vessel (10) and down the annulus (54). At the bottom of the annulus (54) the coolant encounters the flow skirt (50). A major portion of the coolant is then forced through the holes (72) in the flow skirt (50) wherein a portion is directed upward and directly through the lower core support plate (36) while a second portion is directed to the lower portion of the plenum (34) underneath the vortice suppression plate where this portion of the coolant turns upward through the holes in the vortice suppression plate to the lower core support plate. This hydraulic action balances the pressure and flow throughout the underside of the lower core support plate (36). While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention, which is to be given the full breadth of the appended claims and any and all equivalence thereof.
048287903	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS Radionuclides dissolved in the reactor water are incorporated in an oxide film in the course of its formation on the surface of components made of stainless steel by corrosion [e.g., T. Honda et al: Nucl. Technol., 64, 35 (1984)]. According to the study of the present inventors, an oxide film mainly grows in an inner direction (a matrix metal side) at an interface of the oxide film and the matrix metal in high temperature water, and radionuclides transfer by diffusion in the inner direction in the oxide film and then are incorporated in the oxide film at the same interface. The flux (J.sub.0) of radionuclides can be represented by the following equation: ##EQU1## wherein d=the thickness of oxide film k.sub.0 =the constant of proportionality D=the diffusion coefficient C.sub.1 =the concentration of radionuclides in the reactor water C.sub.2 =the concentration of radionuclides at the interface of oxide film/metal Since the thickness of oxide film (d) is a product of the constant of proportionality (k.sub.1) and the amount of the oxide film (m), i.e., EQU d=k.sub.1 m (2) J.sub.o can be represented by the following equation: ##EQU2## On the other hand, the rate of incorporation of radionuclides in the oxide film (J.sub.1) can be represented by the equation (4) using the growth rate of oxide film (dm/dt): ##EQU3## wherein k.sub.2 =the constant of proportionality Since the accumulation rate of radionuclides (J) is J=J.sub.0 =J.sub.1, J can be represented by the equation (5) by eliminating C.sub.2 from the equations (3) and (4): ##EQU4## When the accumulation of radionuclides is rate-determined in the course of diffusion, J can be represented by the following equation: ##EQU5## The equation (6) shows that the accumulation rate (J) is proportional to the diffusion coefficient (D) and means that if the diffusion of radionuclides in the oxide film is inhibited, the accumulation can be inhibited. Therefore, the inhibition of accumulation of radionuclides can be attained by the inhibition of diffusion of radionuclides in the oxide film. This invention is based on such a finding. Major radionuclides contributing to the dose rate are .sup.60 Co and .sup.58 Co, which are present in the cooling water as cations. The oxide surface is hydrolyzed in the solution and charged positively or negatively depending on the pH of the solution as shown in the equations (7) and (8): ##STR1## [see G. A. Parks and P. L. de Bruyn: J. Phys. Chem., 66, 967 (1962)]. Therefore, when the oxide film formed on the component surfaces is positively charged in the cooling water, diffusion of cations of .sup.60 Co and .sup.58 Co in the oxide film can be inhibited, since the oxide film has selective transmission of anions. The pH at electrically neutral state of the oxide surface is defined as a zero point of charge (ZPC). When the pH of the solution is higher than ZPC, the oxide is charged negatively, while when the pH of the solution is lower than ZPC, the oxide is charged positively. Therefore, oxides of ZPC>7 are charged positively in neutral water (pH=about 7) such as cooling water used in a boiling water reactor plant (hereinafter referred to as "BWR plant"). The present inventors have found that when carbon steel, stainless steel, etc. are subjected to an oxidation treatment in a solution containing polyvalent metal cations and anions having a smaller ionic valence number than the cations, for example a solution of Ca(NO.sub.3).sub.2, an oxide film of ZPC>7 can be formed. When such an iron oxide film is formed, the accumulation of radionuclides can be inhibited even if contacted with reactor cooling water. This treating method can be applied whether an iron oxide film is present on the surfaces of components or not. For example, as to stainless steel used in a nuclear power plant in operation, such an object can be attained by pouring a solution containing polyvalent cations and anions having a smaller ionic valence number than the cations into the cooling water. In such a case, the diffusion of cations such as .sup.60 Co, etc. into the oxide film can be inhibited and the accumulation of the cations can also be inhibited. As the polyvalent cations, there can be used at least one member selected from the group consisting of Al.sup.3+, Fe.sup.3+, Ba.sup.2+, Ca.sup.2+, Co.sup.2+, Mg.sup.2+, Ni.sup.2+, Pb.sup.2+, Zn.sup.2+ and Ca.sup.2+. As the anions having a smaller ionic valence number than the cations, there can be used at least one member selected from the group consisting of HCO.sub.3.sup.-, H.sub.2 PO.sub.4.sup.-, MnO.sub.4.sup.-, NO.sub.2.sup.-, NO.sub.3.sup.-, OH.sup.-, HCOO.sup.-, CH.sub.3 COO.sup.-, MoO.sub.4.sup.2-, HPO.sub.4.sup.2-, SO.sub.4.sup.2- and WO.sub.4.sup.2-. The temperature is preferably 150.degree. to 300.degree. C. The concentration of the cations is preferably 3 ppb to 1000 ppm, more preferably 3 to 100 ppb. Usually polyvalent cations as listed in Table 1 are present in the cooling water. TABLE 1 ______________________________________ Maximum concentration Ions (ppb) ______________________________________ Ni.sup.2+ 0.5 Co.sup.2+ 0.05 Zn.sup.2+ 0.5 Cu.sup.2+ 0.5 ______________________________________ [Y. Yuasa: J. Nucl. Sci. Technol., 17, 564 (1980) Therefore, a method of coating the components with an oxide film which can easily adsorb these cations previously is also effective. The present inventors have found that an oxide film formed by treating stainless steel under a weakly oxidizing or reducing atmosphere can satisfy such a condition. The oxide film formed under such conditions have many lattice defects, which become centers of activity and thus show strong adsorbing capacity. As a result, the oxide film is positively charged and inhibit the diffusion of .sup.60 Co and the like into the oxide film by showing selective transmission of anions. The oxidation treatment conditions can be obtained by deaeration so as to make the concentration of dissolved oxygen 10 ppb or less, or the addition of a reducing agent. Examples of the reducing agent are hydrogen, hydrazine, L-ascorbic acid, formaldehyde, oxalic acid, etc. Further, it is also possible to use substances which do not particularly show reducing properties at normal temperatures but can act as a reducing agent at high temperatures. Many organic reagents belong to such substances. That is, organic compounds decompose at high temperatures and special organic compounds act as a reducing agent at such a time. Such special organic compounds are required to be soluble in water and to be decomposed at 300.degree. C. or lower. Further such special organic compounds should not contain elements such as a halogen and sulfur which corrode the matrix such as stainless steel. These elements are possible to cause pinholes and stress cracking by corroding matrix stainless steel. Examples of such organic compounds are organic acids such as oxalic acid, citric acid, acetic acid, formic acid, etc.; chelating agents such as ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), etc. Since these compounds are acidic and very corrosive to the matrix as they are, it is necessary to adjust the pH to 5 to 9 with an alkaline agent such as ammonia, sodium hydroxide, or the like so as to make them neutral or weakly alkaline. Needless to say, salts of these compounds near neutral such as 2-ammonium citrate, EDTA-2NH.sub.4, etc., can be used by simply dissolving them in water. The use of chelating agent such as EDTA, NTA, or the like is particularly preferable, since the chelating agent not only shows reducing properties by decomposition at high temperatures, but also accelerates the dissolution of iron oxide by stabilizing iron ions by chelating so as to finally produce an oxide film having a high chromium content. These organic reducing agents are preferably used in a concentration of 10 ppm to 1% by weight, more preferably 100 to 3000 ppm. If the concentration is too low, no effect is obtained, whereas if the concentration is too high, there takes place incomplete decomposition at high temperatures so as to produce a large amount of sludge which undesirably deposits on piping. The preferable temperature is 150.degree.-300.degree. C. Another method for inhibiting the accumulation of radionculides in the oxide film is to inhibit the incorporation of radionuclides into the oxide film. The radionuclides dissolved in the cooling water is incorporated into the oxide film in the course of its formation on the surface of stainless steel by the corrosion thereof. According to the study of the present inventors, there is the correlation between the deposition rate of radionuclides and the film growth rate. Therefore, it was estimated that the inhibition of film growth resulted in lowering in the deposition. The increase of the film amount (m) of stainless steel under circumstances of cooling water can be represented by a logarithm of time as shown below: EQU m=a log t+b (9) wherein a and b are constants. That is, the growth rate is reduced with the growth of film. Therefore, if a suitable non-radioactive oxide film is formed previously, new formation of film after the immersion in a liquid dissolving radioactive substances can be inhibited. Further, the deposition of radioactive substances taking place at the time of film formation can be inhibited. The present inventors have noticed that the inhibition of deposition of radioactive substances can be attained by previously forming a suitable non-radioactive oxide film on metal components used in contact with the reactor cooling water dissolving the radioactive substances. At the same time, the present inventors have found that the depositiomn rate of .sup.60 Co is dependent on the chromium content in the oxide film previously formed and the deposition rate becomes remarkably small, particularly when the chromium content in the metals constituting the oxide film is 12% by weight or more. Another feature of this invention is based on such a finding. That is, the oxide film previously formed on the surfaces of components contacting with the liquid dissolving radioactive substances contains 12% by weight or more of chromium. By forming the oxide film having such a high chromium content and being positively charged in the reactor cooling water, the deposition of radioactive substances can further be inhibited. The proportion of chromium in the total metals constituting the oxide film (hereinafter referred to as "chromium content") is sufficient when 12% by weight or more. When applied to the BWR plant wherein the cooling water contains about 200 ppb of oxygen, the chromium content in the oxide film gradually decreases due to the oxidation of the chromium in the oxide film to give soluble chromium having a valence number of 6. Therefore, it is desirable to make the chromium content in the oxide film previously formed as high as possible. The oxide film having a chromium content of 12% by weight or more, preferably a remarkably high chromium content, can previously be formed by oxidizing a high chromium content matrix in water at high temperatures, e.g. 150.degree.-300.degree. C. as it is. In the case of carbon steel and low alloy steel, it is difficult to form the oxide film by oxidation in the high temperature water. Further, in the case of 18 Cr--8 Ni stainless steel usually used in nuclear power plants, the chromium content becomes 20% by weight or less when simply oxidized in high temperature water. Therefore, when there is used a raw material which is difficult to form a high chromium content oxide film by simple oxidation in high temperature water, the oxide film having a high chromium content can be formed by covering the surface with a metal coating containing a large amount (about 50% by weight) of chromium, and then oxidizing in water at high temperatures such as 150.degree.-300.degree. C. or in steam at high temperatures such as 150.degree. to 1000.degree. C. The metal coating containing a large amount of chromium can be formed by a conventional method, preferably by a chromium plating method, a chromizing treatment, a chromium vapor deposition method, and the like. On the other hand, when stainless steel is oxidized in water at high temperatures, it is possible to form the oxide film having a chromium content of near 20% by weight. But when such an oxide film is used in the cooling water containing oxygen in the BWR plant mentioned above, the chromium content is gradually lowered due to oxidation to give soluble chromium having a valence number of 6. In such a case, it is desirable to form an oxide film having a higher chromium content previously. This can be attained by carrying out the oxidation in high temperature water containing a reductive substance. The formation of oxide film having such a high chromium content by the above-mentioned method can be explained by the following principle. There are two kinds of oxides of chromium, i.e. chromic oxide (Cr.sub.2 O.sub.3) and chromium trioxide (CrO.sub.3). Chromic oxide is hardly soluble in water, but chromium trioxide is soluble in water. Therefore, oxides of chromium become easily soluble in water under oxidizing circumstances and hardly soluble in water under reducing circumstances. In the case of iron, there are ferrous oxide and ferric oxide. Ferrous oxide is more soluble in water than ferric oxide. Therefore, oxides of iron become more easily soluble in water under reducing circumstances than under oxidizing circumstances. Therefore, when stainless steel containing chromium and iron is oxidized under reducing circumstances, since the iron becomes easily soluble in water and the chromium remains as oxide on the surface of the matrix to form the oxide film having a high chromium content. Even under such reducing circumstances, iron and chromium can be oxidized at high temperatures so long as water is present. The reducing circumstances can be formed by adding a reducing agent to water. Examples of the reducing agent are hydrogen, hydrazine, L-ascorbic acid, formaldehyde, oxalic acid, etc. Further, it is also possible to use substances which do not particularly show reducing properties at normal temperatures but can act as a reducing agent at high temperatures. Many organic reagents belong to such substances. That is, organic compounds decompose at high temperatures and special organic compounds act as a reducing agent at such a time. Such special organic compounds are required to be soluble in water and to be decomposed at 300.degree. C. or lower. Further such special organic compounds should not contain elements such as a halogen and sulfur which corrode the matrix such as stainless steel. These elements are possible to cause pinholes and stress corrosion cracking by corroding matrix stainless steel. Examples of such organic compounds are organic acids such as oxalic acid, citric acid, acetic acid, formic acid, etc.; chelating agents such as ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), etc. Since these compounds are acidic and very corrosive to the matrix as they are, it is necessary to adjust the pH to 5 to 9 with an alkaline agent such as ammonia, sodium hydroxide, or the like so as to make them neutral or weakly alkaline. Needless to say, salts of these compounds near neutral such as 2-ammonium citrate, EDTA-2NH.sub.4, etc., can be used by simply dissolving them in water. The use of chelating agent such as EDTA, NTA, or the like is particularly preferable, since the chelating agent not only shows reducing properties by decomposition at high temperatures, but also accelerates the dissolution of iron oxide by stabilizing iron ions by chelating so as to finally produce an oxide film having a high chromium content. These organic reducing agents are preferably used in a concentration of 10 ppm to 1% by weight, more preferably 100 to 3000 ppm. If the concentration is too low, no effect is obtained, whereas if the concentration is too high, there takes place incomplete decomposition at high temperatures so as to produce a large amount of sludge which undesirably deposits on piping. In the chemical decontamination of nuclear power plants, a decontamination solution containing at least one reagent selected from an organic acid, a chelating agent and a reducing agent is generally used. In order to inhibit a rapid contamination progress after the decontamination, the above-mentioned process is particularly preferable. That is, since the decontamination solution contains the above-mentioned organic compounds, it can be used for the purpose of this invention as it is. But since the decontamination solution after decontamination contains radionuclides such as .sup.60 Co mainly, it cannot be heated as it is due to deposition of .sup.60 Co. Therefore, the abovementioned treatment can be conducted after removing the used decontamination solution, or after removing radionuclides such as .sup.60 Co from the decontamination solution by using a cation exchange resin or electrodeposition, the decontamination solution is heated and the oxide film is formed. When the pH of decontamination solution after decontamination is low, it is adjusted to near neutral by adding an alkaline agent such as ammonium thereto. Further, when the concentration of the organic compounds is too high to conduct the oxidation treatment, a part of the solution is taken out and the solution can be diluted by adding water thereto, or a part of the solution is passed through an ion exchange resin, so as to lower the concentration to the desired value. This invention is illustrated by many of the following Examples, in which all percents are by weight unless otherwise specified. Example 1 Plant component materials made of carbon steel (STPT 42) and stainless steel (SUS 304) having chemical compositions shown in Table 2 were immersed in a cooling water dissolving oxygen in a concentration of 150-170 ppb at a flow rate of 0.5 m/sec at 230.degree. C. for 1000 hours. TABLE 2 ______________________________________ Plant component Chemical composition (%) material Co Ni Cr ______________________________________ STPT 42 0.0063 0.022 0.012 SUS 304 0.22 9.11 18.1 ______________________________________ Then, the resulting oxide films were analyzed by secondary ion mass spectroscopy (SIMS). The results are shown in FIGS. 1 and 2. Distribution of the elements in the thickness direction of oxide film in the case of carbon steel shows that Co, Ni and Cr decrease their concentrations from the surface of the oxide film to the matrix metal. The carbon steel (STPT 42) contains Co, Ni and Cr in very small amounts in the matrix as shown in Table 2, but the contents of these elements in the oxide film are ten to hundred times higher than the original contents as shown in Table 3. Therefore, these elements seem to be incorporated not from the matrix metal but from the cooling water. Further, the oxide film grew at a constant rate with the lapse of time. TABLE 3 ______________________________________ Chemical composition (%) Co Ni Cr ______________________________________ Oxide film 0.0236 1.45 2.95 ______________________________________ More in detail, the oxide film grows to the inner direction at the interface of the oxide film and the matrix metal. On the other hand, the above-mentioned three elements present in the cooling water transmit through the oxide film and reach the above-mentioned interface, and then are incorporated in the growing oxide film. The above-mentioned phenomena can be represented by the following equation; that is, the concentration of ions of elements at the interface of oxide film/metal (C.sub.2) can be represented as follows by using the equations (3) and (4): ##EQU6## When the diffusion coefficient (D) of ions is small and the incorporation of ions in the oxide film is controlled by the diffusion, the equation (10) can be simplified as the following equation: ##EQU7## Therefore, when the growing rate of oxide film (dm/dt) is constant, the concentration of ions of elements at the interface of oxide film and matrix metal (C.sub.2) decreases in order to increase the oxide film amount (m) with the lapse of time; this is in good agreement with the results of SIMS. In the case of stainless steel (SUS 304), concentrations of Ni and Cr in the oxide film are lower than those of the matrix as shown in Table 4. TABLE 4 ______________________________________ Chemical composition (%) Co Ni Cr ______________________________________ Oxide film 0.29 3.07 7.6 ______________________________________ Since Ni and Cr are major elements constituting stainless steel, these elements incorporated in the oxide film seem to be derived from the elements released from the matrix metal by corrosion. FIG. 2 shows a tendency to increase the concentrations of individual elements in the thickness direction of the oxide film. This seems to be that the diffusion of the released elements in the outer direction is prevented by the oxide film, the ion concentrations of these elements at the interface of oxide film/metal increase with the lapse of time, and the oxide film grows at the same interface. As mentioned above, the oxide films of stainless steel and carbon steel clearly grow in the inner direction of the matrix metal in high temperature water. Therefore, radionuclides dissolved in the cooling water seem to transfer in the oxide film by diffusion and to be incorporated in the oxide film at the interface and accumulated. Example 2 Stainless steel (SUS 304) powder and iron powder were subjected to oxidation treatment in a solution of pure water and Ca(NO.sub.3).sub.2 with calcium ion concentration of 50 ppb at 230.degree. C. for 100 hours. FIG. 3 shows the results of zeta potential of stainless steel powder after the oxidation treatment and FIG. 4 shows those of iron powder after the oxidation treatment. Table 5 shows ZPC of individual oxides. TABLE 5 ______________________________________ Powder Treating conditions ZPC ______________________________________ Stainless In pure water 7 steel In aq. solution of 11 Ca(NO.sub.3).sub.2 (Ca.sup.2+, 50 ppb) Iron In pure water 7 In aq. solution of 11.5 Ca(NO.sub.3).sub.2 (Ca.sup.2+, 50 ppb) ______________________________________ As is clear from Table 5, when stainless steel and iron are subjected to the oxidation treatment in pure water, ZPC is 7 in each case, while when subjected to the oxidation treatment in the aqueous solution of Ca(NO.sub.3).sub.2, ZPC is 11 in the case of stainless steel and 11.5 in the case of iron, and the resulting oxidized products are charged positively in neutral water (pH 7). Therefore, when subjected to the oxidation treatment in a solution containing a combination of divalent cation Ca.sup.2+ and monovalent anion NO.sub.3.sup.- (i.e. in Ca(NO.sub.3).sub.2 solution), it becomes clear that the oxide film is charged positively in neutral water, shows anion selective transmission, and inhibits transmission of cations such as .sup.60 Co in the cooling water. The combination of a polyvalent metal cation and an anion having a lower valence number than the cation can be selected optionally. But considering problems of corrosion of materials such as stress cracking by corrosion, toxicity, etc., the combination I or II shown in Table 6 is preferable. TABLE 6 ______________________________________ Combi- nation Polyvalent metal cation Anion ______________________________________ I Al.sup.3+, Fe.sup.3+, Ba.sup.2+, Ca.sup.2+, HCO.sub.3.sup.-, H.sub.2 PO.sub.4.sup.-, Co.sup.2+, Mg.sup.2+, Ni.sup.2+, Pb.sup.2+, MnO.sub.4.sup.-, NO.sub.2.sup.-, NO.sub.3.sup.-, Zn.sup.2+, Cu.sup.2+ OH.sup.-, HCOO.sup.-, CH.sub.3 COO.sup.- II Al.sup.3+, Fe.sup.3+ MoO.sub.4.sup.2-, HPO.sub.4.sup.2-, SO.sub.4.sup.2-, WO.sub.4.sup.2- ______________________________________ The concentrations of these ions are not critical and can be usable up to the saturated solubility of chemical substances mentioned above. But when the concentrations are too high, there arises a problem of corrosion of the material. Therefore, the concentration of 3 ppm to 1000 ppm is generally preferable. The temperature for the oxidation treatment is preferably 150.degree. C. or higher, more preferably 200.degree. to 300.degree. C., since too low temperature for the oxidation treatment takes a longer time for the growth of oxide film. The thickness of the oxide film is preferably 300 .ANG. or more. Example 3 Stainless steel (SUS 304) powder and iron powder were subjected to oxidation treatment in deaerated neutral pure water at 288.degree. C. for 100 hours. Then, zeta potentials of the thus treated materials were measured in a KNO.sub.3 solution (0.01M, outside of this invention), or in nitrate solutions of Co.sup.2+, Ni.sup.2+, and Zn.sup.2+ in concentrations of 50 ppb as divalent cations. The results are shown in FIGS. 5 and 6. X-ray diffraction of the resulting oxide films formed on the surfaces of stainless steel and iron revealed that they were magnitude (Fe.sub.3 O.sub.4). In each case, the zeta potential transfered to the positive direction in the presence of polyvalent metal cations and took the positive value in neutral water. Example 4 After immersing stainless steel having a chemical composition as shown in Table 7 in the cooling water flowing at a rate of 0.5 m/sec for 1000 hours at 230.degree. C., the amount of oxide film and the deposited .sup.60 Co amount were measured. TABLE 7 ______________________________________ Chemical composition (%) C Si Mn S Ni Cr Co P ______________________________________ SUS 304 0.06 0.76 1.12 0.023 9.11 18.07 0.22 0.029 ______________________________________ Before the immersion, the stainless steel was subjected to mechanical processing on the surface, degreasing and washing. The cooling water continued .sup.60 Co in a concentration of 1.times.10.sup.-4 .mu.Ci/ml and 90% or more of .sup.60 Co was present as ions, dissolved oxygen in a concentration of 150-170 ppb, and had a temperature of 230.degree. C. and a pH of 6.9-72. In this Example, the stainless steel was subjected to oxidation treatment by immersing it in flowing pure water at 285.degree. C. having a dissolved oxygen concentration of 200 ppb or less and an electrical conductivity of 0.1 .mu.S/cm for 50 to 500 hours to previously form an oxide film having a chromium content of 12% or more. FIG. 7 shows the change of amount of typical elements in the oxide film (as a total of Fe, Co, Ni and Cr) with the lapse of time. As is clear from FIG. 7, the amount increases according to a rule of logarithm after 100 hours. FIG. 8 shows the amount of .sup.60 Co deposited with the lapse of time. As is clear from FIG. 8, the amount also increases according to a rule of logarithm after 100 hours as in the case of FIG. 7. Therefore, FIGS. 7 and 8 clearly show that the deposition rate of .sup.60 Co is rate-determined by the oxide film growth rate. Further, the growth rate of oxide film becomes smaller with the progress of growth. Example 5 On the surface of the same stainless steel as used in Example 4, non-radioactive oxide films having a chromium content of 5.2 to 20.3% in the total metal elements were previously formed, respectively. Individual oxide films were immersed in the cooling water under the same conditions as described in Example 4 to measure the deposition rate of .sup.60 Co. The results are shown in Table 8 and FIG. 9. TABLE 8 ______________________________________ Composition of Deposition oxide film (%) rate of .sup.60 Co Run No. Cr Ni Fe (.mu.Ci/cm.sup.2 .multidot. hr) ______________________________________ 1 5.2 4.9 89.9 0.27/t 2 6.6 3.0 90.4 0.27/t 3 7.9 2.8 89.4 0.27/t 4 10.1 6.4 83.5 0.27/t 5 12.0 4.0 84.0 0.0562/t 6 20.3 4.7 75.0 0.0984/t ______________________________________ In Table 8, t is a total time in hour of the preoxidation treatment time and the immersion time in the cooling water. FIG. 9 shows the amount of oxide film formed when the stainless steel is subjected to oxidation treatment at 130.degree. to 280.degree. C. for 6000 hours. As is clear from FIG. 9, the formation of oxide film is accelerated at 150.degree. C. or higher with an increase of the temperature, and particularly remarkably over 200.degree. C. Therefore, the oxidation treatment temperature is particularly preferable over 200.degree. C. The reactor water temperature in an operating BWR plant is 288.degree. C., and the effective oxide film can be formed at such a temperature. As is clear from Table 8 and FIG. 10, the deposition rate of .sup.60 Co (dS/dt) is in inverse proportion to a total time (t) of the time required for previous oxidation treatment (the pre-oxidation treatment time, t.sub.0) and the immersion time in the cooling water (t.sub.1), and can be represented by the following equation in each case: ##EQU8## wherein k is a constant depending on the kind of oxide film formed by the pre-oxidation treatment, and conditions such as .sup.60 Co concentration in the solution dissolving radionuclides, temperatures, etc. Therefore, in order to make the deposition rate of .sup.60 Co small after immersion in the solution dissolving radionuclides under constant conditions, the pre-oxidation treatment time (t.sub.0) is made larger, or alternatively proper pre-oxidation treatment conditions are selected so as to make the constant k smaller. But to make the pre-oxidation treatment time (t.sub.0) larger is not advantageous from an industrial point of view, it is desirable to select an oxide film having a chromium content of 12% or more so as to make the constant k smaller and to reduce the deposition rate of .sup.60 Co. Example 6 The same stainless steel as used in Example 4 was held in water containing a reducing agent as listed in Table 9 in an amount of 1000 ppm at 250.degree. C. for 300 hours. The pH of water was adjusted to 7 with ammonia. The resulting oxide film formed on the surface of stainless steel was peeled off in an iodine-methanol solution and the chromium content in the oxide film was measured by conventional chemical analysis. The results are shown in Table 9. As is clear from Table 9, oxide films having a very high chromium content were able to be obtained by the addition of a reducing agent. Particularly, the addition of a chelating agent such as Ni salt of EDTA or Ni salt of NTA makes the chromium content remarkably high. TABLE 9 ______________________________________ Cr Content in Run No. Reducing agent oxide film (%) ______________________________________ 1 None (pure water) 19 2 Hydrazine 30 3 Oxalic acid 32 4 Citric acid 35 5 EDTA--Ni 63 6 NTA--Ni 55 7 Hydrogen (saturated) 28 ______________________________________ Example 7 The same stainless steel as used in Example 4 was held in water containing 1000 ppm of EDTA at a temperature of 100.degree. to 300.degree. C. for 300 hours. The chromium content in the resulting oxide film was measured in the same manner as described in Example 6. The results are shown in Table 10. TABLE 10 ______________________________________ Temperature Cr Content in Run No. (.degree.C.) oxide film (%) ______________________________________ 1 100 No oxide film was formed. 2 150 70 3 200 79 4 250 63 5 300 58 ______________________________________ As is clear from Table 10, when the temperature is 100.degree. C. or lower, no oxide film is formed, so that the oxidation treatment is preferably conducted at 150.degree. C. or higher. Example 8 Stainless steel (SUS 304) the surface of which had been polished was subjected to oxidation treatment previously under the conditions as shown in Table 11. Then, the thus treated stainless steel was immersed in a CoSO.sub.4 solution containing 50 ppb of Co.sup.2+ ions at 285.degree. C. (the same temperature as that of cooling water in a BWR plant) for 200 hours. The deposited Co amount was measured. TABLE 11 ______________________________________ Run Temperature Time No Solution Concentration (.degree.C.) (hrs) ______________________________________ 1* Ca(NO.sub.3).sub.2 Ca.sup.2+ : 50 ppb 230 100 2* EDTA--Ni 1000 ppm 230 100 3* NTA--Ni 1000 ppm 230 100 4 CaSO.sub.4 Ca.sup.2+ : 50 ppb 230 100 5 Pure water 230 100 (O.sub.2 : 200 ppb) 6 No oxidation treatment was conducted. ______________________________________ Note This invention The deposited amount of cobalt was evaluated by using an energy dispersing type X-ray analyzer (EDX) and obtaining Co/Fe ratios by dividing the peak strength of Co by the peak strength of Fe. The results are shown in Table 12. TABLE 12 ______________________________________ Run No. Co/Fe ratio ______________________________________ 1 <0.1 2* 0.1 3* 0.1 4 0.5 5 0.6 6 1.5 ______________________________________ Note *This invention As is clear from Tables 11 and 12, when the oxidation treatments were conducted as shown in Run Nos. 4 and 5, the deposited cobalt amount could be reduced to about 1/3 of that of Run No. 6 wherein no oxidation treatment was conducted, but the inhibition effect is not sufficient. In contrast, when the oxidation treatment was conducted as shown in Run Nos. 1 to 3 which belong to this invention, the deposition of cobalt was inhibited remarkably effectively. In addition, when the oxidation treatment is conducted by using the solutions of Run Nos. 1 and 2 or Run Nos. 1 and 3, the more effective inhibition can be expected. This invention can be applied to nuclear power plants as follows. (1) In the case of re-use of piping and devices used in nuclear power plants after decontamination by the chemical method and the like, since the oxide film on the surfaces of components is dissolved and peeled off by the decontamination operation, the metal base is exposed and the depositing amount of radionuclides at the time of re-use shows the same change with the lapse of time as shown in FIG. 8. In such a case, when the oxidation treatment of this invention is applied before the re-use, the deposition of radioactive substances can be inhibited. (2) This invention can be applied to any kinds of nuclear power plants. For example, in the case of BWR plant, a pressure vessel, re-circulation system piping and primary cooling water cleaning system piping, etc., contact with reactor water containing radioactive substances; and in the case of a pressurized water type nuclear power plant, a pressure vessel, components in a reactor, a vapor generator, etc., contact with the same reactor water as mentioned above. Therefore, by applying this invention to the whole or a part of components made of at least one metal selected from stainless steel, Inconel, carbon steel and, Stellite, the deposition of radioactive substances on the surfaces of components can be inhibited and it becomes possible to provide nuclear power plants wherein workers are by far less exposed to radioactive irradiation. (3) The oxide film can be formed by this invention on surfaces of components contacting with the cooling water dissolving radioactive substances before or after the construction of nuclear power plants. The oxidation treatment after enrichment of chromium content in the surface portion of the base metal can be conducted either before the construction of the plants, or after construction of the plants by introducing high-temperature water or hight-temperature steam. (4) To already constructed plant piping and devices, this invention can be applied as follows. (a) In the case of a BWR plant as shown in FIG. 11, the solutions of compounds as shown in Example 2 or 6 can be poured into the primary cooling water using a pouring apparatus. In FIG. 11, numeral 1 denotes a reactor, numeral 2 a turbine, numeral 3 a hot well, numeral 4 a low pressure condensed water pump, numeral 5 a demineralizer for condensed water, numerals 6a and 6b are the abovementioned pouring apparatus, numerals 7a and 7b are dissolved oxygen concentration meters, numeral 8a supplying water heater, numeral 9 a demineralizer for reactor cleaning system, and numeral 10 a recirculation system. In this invention, the pouring apparatus can be attached to, for example, a down stream of the demineralizer for condensed water (5) in the condensed water system and/or a down stream of the supplying water heater (8) in the water supplying system. The pouring amount can be controlled by sampling the reactor water and measuring the concentration of polyvalent cations or oxygen concentration. Further, the cooling water can be sampled preferably at a position of inlet for reactor water cleaning 3. PA1 (b) The pouring of polyvalent metal cations can be replaced by placing a metal which can release polyvalent metal cations in a solution. For example, a zinc, magnesium or aluminum plate is placed as a sacrificial anode in a condensate hot well 4 shown in FIG. 11. By this, Zn.sup.2+, Mg.sup.2+, or Al.sup.3+ ions are released in the primary cooling water to increase the polyvalent metal cation concentration in the cooling water system and to obtain the same effect as obtained in (a) mention above. Further, this is also effective for preventing corrosion of the hot well 4. It is also effective to attach an alloy filter containing zinc, aluminum, etc., to a condensate cleaning system 5 or a cooling water cleaning system 6 shown in FIG. 11. By this, the same effect as obtained in (a) mentioned above as well as crud removing effect can be obtained.
051538989	abstract	An X-ray exposure apparatus includes a stage for holding a mask having a pattern for circuit manufacturing, a stage for holding a wafer to be exposed to the pattern of the mask with X-rays, and a reflection reduction imaging system, disposed between the mask stage and the wafer stage, including a reflecting mirror arrangement, containing at least three, but not more than five, reflecting mirrors coated with multi-layer films for receiving X-rays from the mask and directing them to the wafer to expose the wafer to the pattern of the mask with the X-ray in a reduced scale.
044341300	summary	TECHNICAL FIELD This invention relates to the generation of energy from fusions of atomic nuclei which are caused to travel towards each other along collision courses through a compressed sheath of spiraling electrons. BACKGROUND ART It is known that individual nuclear particles are so constituted as to permit fusing of the lighter nuclei. Fusion of light nuclei is accompanied by release of energy. Of particular interest is any fusion reaction in which power can be produced in quantities greater than the power consumed in establishing and maintaining the reaction. There are over 30 reactions now known to be possible. The most appealing reactions are those which involve the heavy hydrogen isotopes, deuterium and tritium, because they tend to have the largest fusion reaction cross section at the lowest energies. Many possible reactions are well known. For example, Van Nostrand's Scientific Encyclopedia, Fifth Edition, Reinhold Company, New York, N.Y., 1976, at page 1656, et seq., discusses various possibilities for producing a net gain in power from fusion reactions and briefly describes some of the attempts to perform such reactions with a net power gain. Plasma research has received and is receiving concentrated attention, but the formidable task of plasma containment has yet to be solved. In avoidance of the problems of containment, a more recent approach involves laser-induced fusion. In its simplest form a focused energetic laser beam is brought to bear on a small deuterium-tritium pellet for heating to fusion temperatures. Efforts on this and on other fronts, such as those involving containment, have been steady in response to high incentives. Thus, while many of the possibilities have long been known and have been widely attached through various approaches towards achieving net power gain from fusion, the challenge remains unsatisfied. The obvious advantage of fusion power is that it offers the promise of being able to utilize an essentially inexhaustible low cost fuel supply. This prospect is a growing challenge as world demands for energy continue to increase. A further significant advantage is that optimum fuels may be chosen to produce reaction products which are completely non-toxic and thus permit energy producing operations compatible with the most demanding environmental requirements. Reactions free of neutron generation can produce energy in a way that is shieldable for personnel protection simply by the presence of structures necessary for carrying out the fusion reaction. In application of this invention, the individual reactants are combined in such a way that they are not individually self-reactant. This permits definitive choice and execution of neutronless fusion. DISCLOSURE OF THE INVENTION The present invention provides a new method and system of atomic fusion for power generation. It is based upon the control of oppositely directed ions by a spiral electron sheath as to promote fusion. Appropriately high velocity fusible ion beams are directed along head-on collision paths in an annular zone wherein beam compression by electrostatic focusing of a sheath of spiraling electrons forces the ions to follow a path where the potential gradient is minimum. The invention involves generating two beams of fusible ions which are projected towards each other along cylindrical paths determined by the electrical field produced by a compressed spiral sheath of electrons traversing a cylindrical reaction zone. In one embodiment, a steady radial electric field is imposed on an electron beam to compress the beam and reduce the radius of the spiral paths for enhancing electron density. Ions of one beam travel the same cylindrical zone as ions of the other beam. Energy produced by the nuclear reactions resulting from such collisions is then extracted by means of heat exchange or deceleration processes. Beam compression to concentrate the electrons is achieved through electrostatic focusing. The electron distribution across the electron sheath will not be uniform and will exhibit a zone minimum gradient which defines the location of the path the ions will be forced to follow. Thus, two independent streams of ions can be propelled in opposite directions in the same space to maximize fusion-producing collisions. Energy production involving the present invention, through choice of fuels, can be free of hazards of undesirable radiation and avoid the production of toxic wastes that have characterized operations involving fission as practiced. Because of this characteristic of the present invention, it lends itself particularly to reactors of the size of drive systems for automobiles, aircraft or smaller vehicles, as well as satisfying the utility needs for living quarters and for industry. As will be hereinafter noted, several fusion reactions are known which involve the use of fuels which are plentiful and available throughout the world. The isotope deuterium is plentiful in the sea. Helium-3, boron and lithium are known to be in such supply as to warrant labelling them and deuterium as fuels which are essentially inexhaustible. By directing ions of such fuel at appropriate energies along collision paths, a measure of control is provided in energy production from fusion reactions that has heretofore not appeared possible. With the foregoing points in mind, embodiments of the invention and unique systems employed for carrying out fusion reactions will now be described.
041845145	claims	1. A valve system utilizing three independent input signals combined in a two-out-of-three logic function for actuation of de-actuation of a normally pressurized associated mechanism operatively connected to a pressure supply and a vent, comprising: valve means composed of a plurality of valve assemblies positioned intermediate an associated pressure supply, a vent, and an associated mechanism to be actuated or de-actuated; a first pair of said valve assemblies being positioned intermediate an associated pressure supply and a vent, a second pair of said valve assemblies being positioned intermediate a vent and an associated mechanism to be actuated or de-actuated; valve actuation/de-actuation signals a, b, and c, for actuation/de-actuation of said valve means; and means for initiating said signals a, b, and c; whereby failure of one of signals a, b, and c, or failure of a corresponding valve assembly will neither prevent nor cause action of an associated mechanism. 2. The valve system defined in claim 1, wherein another of said valve assemblies is in common with one of said first pair of valve assemblies positioned intermediate an associated pressure supply and said vent, and having an ouput therefrom connected intermediate said second pair of valve assemblies positioned intermediate said vent and an associated mechanism. 3. The valve system defined in claim 1, wherein another of said valve assemblies is positioned so as to have an input connected intermediate said first pair of valve assemblies and an output connected intermediate said second pair of valve assemblies. 4. The valve system defined in claim 1, wherein another of said valve assemblies is positioned so as to have an input connected intermediate said first pair of valve assemblies and an output connected to an output of one of said first pair of valve assemblies. 5. The valve system defined in claim 1, wherein said actuation/de-actuation signals define logic symbols a,b,c and a,b,c and conditions of actuation and de-actuation are expressed by Boolean equations ab+bc+ca+abc, ab+bc+ca+abc, and equivalent variations thereof.
	description	The present application claims priority from Japanese application JP 2003-398781 filed on Nov. 28, 2003, the content of which is hereby incorporated by reference into this application. The present invention relates generally to an X-ray imaging apparatus and an X-ray imaging method, and more particularly to an apparatus and a method for nondestructively inspecting the inside of an object. As an imaging apparatus for nondestructively observing the inside of a sample by using X-rays, there are an absorption contrast X-ray imaging apparatus using changes in an intensity caused by a sample as contrast, and a phase contrast X-ray imaging apparatus using phase-shift caused by a sample as contrast. The former absorption contrast X-ray imaging apparatus mainly comprises an X-ray source, a sample positioning mechanism and a detector. The X-rays emitted from the X-ray source are irradiated on a sample positioned by the sample positioning mechanism, and detected by the detector after passing through the sample. Thus an image constituting the changes in the intensity caused by absorption of the sample as contrast was obtained. Since a measuring principle and a structure of the apparatus are relatively simple, this imaging technique is widely used in many fields including a medical diagnostic field. In the case of two dimensional observation, it is called roentgen, and in the case of three dimensional observation by a Computed Tomography (CT) scanner, it was called X-ray CT. On the other hand, there are an apparatus described in the Non-Patent Document 1 (Appl. Phys. Lett. 6, 155 (1965)), an apparatus disclosed in the Patent Publication Document 1 (JP-A-10-248833) and the like as the latter phase contrast X-ray imaging apparatus. In general, the phase-shift caused by a sample is remarkable large as compared with changes in the intensity, so the phase contrast X-ray imaging apparatus has an advantage in that the sensitivity is higher than that of the absorption contrast X-ray imaging apparatus. For this reason, even if an imaging target is biological soft tissue composed mainly of light elements such as oxygen and carbon, which have been difficult to observe by the absorption contrast X-ray imaging apparatus, the phase contrast X-ray imaging apparatus makes it possible to observe the inside structure of it with low X-ray exposures and without using contrast agents nondestructively and in a state of high-sensitivity. The phase contrast X-ray imaging apparatus as described above is constituted by adding an X-ray interferometer such as a Bonse-Hart interferometer (described in Non-Patent Document 1 (Appl. Phys. Lett. 6, 155 (1965)), or an interferometer (described in Non-Patent Document 2 (J. Appl. Cryst. 7, 593 (1974)) which is composed by two crystal blocks, to an X-ray source, a sample positioning mechanism and a detector. The Bonse-Hart interferometer, as shown in FIG. 1, has three wafers (beam splitter 1, mirror 2 and analyzer 3) arranged in parallel and at equal intervals on a crystal block fabricated from a perfect crystal ingot. An incident X-ray 4 is split into two beams such as a beam 5 and a beam 6 by the beam splitter 1, the beams are reflected by the mirror 2 and are combined on the analyzer 3 to form two interference beams 7 and 8. When a sample 9 is positioned in an optical path of the beam 5 or the beam 6, the phase shift caused by the sample 9 appears as the changes in the intensities of interference beams 7 and 8 by the principle of superposition (interference pattern). The phase map (spatial distribution of phase-shift caused by the sample) was calculated from the interference patterns detected by an image detector. Furthermore, an imaging apparatus enabling the three dimensional nondestructive observation by combining the phase contrast imaging method and the method for normal X-ray CT, is disclosed in Patent Publication Document 2 (JP-A-4-348262) etc. In this case, the X-ray is irradiated on the sample from a plurality of directions different from each other in the same way as that of the normal X-ray CT, and a cross sectional image of the sample is reconstructed from the phase contrast projection images obtained for respective projections. The X-ray is approximately transparent for the light elements such as oxygen and carbon, and almost all of incident X-rays are passed through the object. Therefore, the change in the intensity caused by absorption of a sample is extremely small, and it is difficult to perform the fine observations of a sample mainly composed by the light elements such as biological soft tissues and organic material etc. by the absorption contrast X-ray imaging apparatus. In order to improve the sensitivity, the contrast agents and/or an extension of exposure time is tried to be used. However, in this case, some problems that a position can be observed is limited and/or an X-ray exposure is increased, has been occurred. On the other hand, though the sensitivity of the phase contrast X-ray imaging method is satisfactorily sufficient, the phase-shift α generated by the sample is detected as a wrapped value α′ (α′=α−Int(α/2π)2π) in its region of 0−2π, as shown in FIG. 2. Therefore, a complicated calculation called a phase unwrapping method (described in JP-A-2001-153797) is required to obtain the true phase-shift α. Furthermore, in a region where a shape and an internal structure of the sample is complicated and its density is rapidly varies spatially, an optical path deviates from an original optical path by X-ray diffraction, the visibility of an interference pattern is lowered or an interference fringes are disappeared. As a result, the unwrapping processing cannot be performed normally and the phase-shift α cannot be obtained accurately. To avoid this problem, a method that the sample is placed in a sample cell filled liquid to reduce a difference of density between the sample and its periphery, is disclosed in JP-A-7-209212. An influence of the shape of the sample can be reduced by this method, but, the influence of the rapid change in the density inside the sample is not avoidable. As evident from the above description, a sensitivity range of conventional absorption X-ray imaging and a sensitivity range of the phase contrast X-ray imaging are separated into both ends as shown in FIG. 3. For the sample including an organ of large density change such as the bone or the lung and an organ of small density change such as biological soft tissue, neither of the imaging methods can observe the sample with satisfactory sensitivity. The purpose of the present invention is to provide an imaging apparatus and imaging method having a wide sensitivity range as shown FIG. 3, and enabling observations of a sample including an organ of large density change such as the bone or the lung and an organ of small density change such as biological soft tissue. FIG. 4 shows a functional diagram of general X-ray interferometer. As shown in FIG. 4, an incident X-ray beam 14 is split into a first beam 15 and a second beam 16. These beams are reflected by a mirror optical system 12 respectively, and combined by an analyzer 13 to form an interference beam 17. The intensity I of the interference beam 17 is given by Equation (1),I=I1+I2+2r√{square root over (I1·I2)}·cos(φ) (1)where I1 is the intensities of the beam 15, I2 is the intensities of beam 16, r is a degree of coherence, φ is a phase difference between the beam 15 and 16. The r is given by Equation (2), r = 2 ⁢ J ⁡ ( v ) v ( 2 ) v = 2 ⁢ π ⁢ W ⁢ ⁢ Δ ⁢ ⁢ x λ ( 3 ) where λ is a wavelength of an incident X-ray beam, W is a divergence angle of the incident X-ray, Δx is deviation distance between the beam 15 and 16 on the analyzer 13, and J is Bessel function of the first kind. If the X-ray interferometer is fabricated as designed, the optical paths of the beam 15 and 16 on the analyzer 13 are completely coincident with each other (Δx=0), and r becomes 1. On the other hand, if the X-ray interferometer has deformation of crystal block, the optical paths of the beam 15 and 16 deviate from each other on the analyzer 13 (Δx≠0), and r is reduced with increasing of Δx. Hereinafter, in order to simplify the calculation, it is assumed that Δx=0. A sample 18 which generates the change in the intensity ΔI, phase-shift Δp, and the change in the direction of the X-ray Δθ, is placed in an optical path of the first beam 15 as shown FIG. 4, the Equation (1) becomes Equation (4),I′=(I1−ΔI)+I2+2r′√{square root over ((I1−ΔI)·I2)}·cos(ø+Δp) (4)where r′ is a changed degree of coherence caused by Δθ. Assuming that a distance between the sample 18 and the analyzer 13 is R, the deviation distance Δx′ is given by Equation (5).Δx′=RΔθ (5)Accordingly, r′ is expressed by Equation (6) as a function of Δθ from Equation (2) and Equation (3), r ′ = 2 ⁢ J ⁡ ( kWR ⁢ ⁢ Δ ⁢ ⁢ θ ) kWR ⁢ ⁢ Δ ⁢ ⁢ θ ⁢ ⁢ where ⁢ ⁢ k = 2 ⁢ π ⁢ 1 λ . ( 6 ) Since the Δθ is given by spatial phase-shift incline of the sample (see FIG. 5), Δ ⁢ ⁢ θ = Δ ⁢ ⁢ p Δ ⁢ ⁢ s ( 7 ) Equation (6) can be expressed by r ′ = 2 ⁢ J ⁡ ( WR ⁢ Δ ⁢ ⁢ p Δ ⁢ ⁢ s ) WR ⁢ Δ ⁢ ⁢ p Δ ⁢ ⁢ s . ( 8 ) Furthermore, Equation (8) can be approximated as a linear function in a range 0<WRΔp/Δs<1 as shown in Equation (9). Eventually, it can be considered that r′ is approximately proportional to the spatial phase-shift incline of the sample (Δp/Δs). r ′ = 1 - WR ⁢ ⁢ Δ ⁢ ⁢ p 3.8 ⁢ ⁢ Δ ⁢ ⁢ s ( 9 ) Next, the sensitivity of the imaging method according to the present invention and the conventional imaging methods are compared using the above equations. A sample having a cross sectional shape as shown in FIG. 5 (a thickness varies linearly from t1 to t2 with a width Δs at a central portion) and uniform density is used for this comparison. Using the complex refractive index n of the samplen=1−δ+iβ (10),a relative intensity difference ΔI between IA and IB which passed through a region A and a region B respectively, is given by I A - I B I = Δ ⁢ ⁢ I I = 4 ⁢ π ⁡ ( t 1 - t 2 ) ⁢ β λ ( 11 ) On the other hand, a phase-shift difference Δp (=pA−pB) is expressed by Equation (12). p ⁢ ⁢ A - p ⁢ ⁢ B = Δ ⁢ ⁢ p = 2 ⁢ π ⁡ ( t 1 - t 2 ) ⁢ δ λ ( 12 ) Using Equation (12), the spatial phase-shift incline (Δp/Δs) at a boundary of the region A and the region B is expressed by Equation (13). Δ ⁢ ⁢ p Δ ⁢ ⁢ s = 2 ⁢ π ⁡ ( t 1 - t 2 ) ⁢ δ λ ( 13 ) Therefore, a relative change ΔA of an amplitude A (=2r(I1·I2)1/2) of the interference beam is given by Equation (14) from Equation (1) and Equation (9) with assuming IA≈IB. Δ ⁢ ⁢ A A = 2 ⁢ π ⁡ ( t 1 - t 2 ) ⁢ δ ⁢ WR λ ⁢ ⁢ Δ ⁢ ⁢ s ( 14 ) The ΔA is proportional to the spatial phase-shift incline (Δp/Δs) and becomes 0 excepting the boundary regions, so that ΔA cannot be simply compared with ΔI/I and Δp. Accordingly, the sensitivities of the respective imaging methods are compared by a relative signal amount at the boundary region. From Equation (12) and Equation (14), a ratio of the sensitivity of the phase contrast X-ray imaging method and the present invention is expressed by Equation (15). Δ ⁢ ⁢ p ⁢ : ⁢ Δ ⁢ ⁢ A A = 1 ⁢ : ⁢ WR Δ ⁢ ⁢ s ( 15 ) From Equation (11) and Equation (14), the ratio of the sensitivity of the imaging method according to the present invention and the absorption contrast X-ray imaging method is expressed by Equation (16). Δ ⁢ ⁢ A A : Δ ⁢ ⁢ I I = δ · WR Δ ⁢ ⁢ s : 2 ⁢ ⁢ β ( 16 ) In an imaging using the X-ray interferometer made of a perfect silicon crystal, the divergence angle W of the incident X-ray is about the width (W≈105) of Bragg diffraction of the crystal, and the distance R between the sample and the analyzer is approximately 10 cm. Furthermore, the minimum value of detectable Δs is determined approximately by spatial resolution of the imaging detector (˜10−6 m). Thus Δp>ΔA/A is obtained from Equation (15). In addition, since α<<β for the light elements, ΔA/A>ΔI/I is obtained from Equation (16). To sum up the above consideration, the sensitivity range of the imaging method using the change in the amplitude ΔA/A as contrast (the present invention) can be drawn at the middle between the phase contrast X-ray imaging method and the absorption contrast X-ray imaging method as shown in FIG. 3. Moreover, the amplitude A of the interference beam is proportional to the intensity I of the interference beam (Equation (1) and Equation (4)), and therefore A can be directly calculated from 1 without executing special operational calculation such as unwrapping processing. Accordingly, the imaging method using the change of amplitude ΔA/A as the contrast (the present invention) can cover widely the sensitivity range of the absorption contrast X-ray imaging method. Furthermore, Δp and ΔA/A are measured simultaneously by a Fourier transformation method or a fringe scanning method. As a result, by combining Δp and ΔA/A, all of the sensitivity range can be covered as shown in FIG. 3. As explained above, aforementioned problem has been solved by the imaging apparatus and imaging method using the amount of change in the amplitude as the contrast, and/or the amount composed of the amount of change in the amplitude and the amount of phase-shift as the contrast. Hereinafter, embodiments of the present invention will be described with reference to the drawings. In the drawings, elements having the same function identified by the same reference designation. FIG. 6 is a view showing a constitution of one example of an X-ray imaging apparatus used in the present invention. The present X-ray imaging apparatus comprises an X-ray interferometer 19, a position adjusting mechanism 20 for an X-ray interferometer, a sample holder 21, a sample holder positioning mechanism 22, a phase shifter 23, a phase shifter positioning mechanism 24, an X-ray detector 25, a controller 26, a processing portion 27 and a display unit 28. In this embodiment, a Bonse-Hart interferometer shown in FIG. 1 is used as the interferometer 19. An incident X-ray 29 is split, reflected and combined sequentially by a splitter 30, a mirror 31, and an analyzer 32, to from a first interference beam 32a and a second interference beam 32b in this interferometer. When a sample is placed in an optical path of split one beam by using a sample holder 21 positioned by a sample holder positioning mechanism 22, X-ray intensity, phase and a transmission direction of an X-ray are changed by the sample. As a result, the intensity of the interference beams 32a and 32b are changed, a change in an amplitude and a phase caused by the sample can be obtained by means of a fringe scanning method shown below. If a two dimensional X-ray detector is used as the X-ray detector 25, the above described change can be detected as a two dimensional image. In the embodiment 1, measurement is carried out according to a flow shown in FIG. 7. In order to exclude a background distribution (a spatial distribution with no sample) of the amplitude and the phase caused by a crystal distortion etc. of the X-ray interferometer 19, an image is obtained by measuring according to the following procedure. (1) Before placing a sample, a background distribution of an amplitude (A0) and a phase (Δp0) are obtained by using the fringe scanning method (step 701—measurement of background data). (2) The sample is placed in the optical path by using the sample holder 21 and the sample holder positioning mechanism 22 (step 702—installation of sample). (3) A distribution of an amplitude (Al) and a phase (Δp1) including the background and the sample is obtained by means of the fringe scanning method (step 703—actual measurement). (4) The change in the amplitude ΔA (=A1/A0) caused by the sample is calculated (step 704—calculation of ΔA image) from a distributed image of the amplitude obtained at the above described steps (1) and (3). After performing above procedure, the obtained ΔA image is displayed at a display unit 28 (step 708). As mentioned before, a phase-shift Δp can simultaneously be measured by the fringe scanning method and Fourier transformation method. Accordingly, by combining ΔA/A and Δp, all of the sensitivity range can be covered as shown in FIG. 3. Following the processing in FIG. 7, in FIG. 8, (5) From the distribution image of the phase-shift obtained by the above described steps (1) and (3), the phase-shift Δp (=Δp1−Δp0) caused by the sample is calculated (step 704′—calculation of Δp image) with conventional unwrapping processing. Incidentally, the fringe scanning method is performed by changing the phase of interfering beams. The phase is shifted by rotating or moving the phase shifter 23 in parallel with respect to the X-ray by using the phase shifter positioning mechanism 24. The changes in the amplitude and the phase are calculated from the obtained interference images at each phase-shift. In case of M sheets of the interference images whose phases are shifted at equal intervals, the amplitude A is given by an Equation (17). A = Abs ⁡ [ ∑ m = 0 M - 1 ⁢ I m ⁢ ⁢ exp ⁡ ( - 2 ⁢ ⁢ π ⁢ ⁢ ⅈ ⁢ ⁢ m M ) ] ( 17 ) where Im is an interference image with phase difference 2πm/M. In addition, a phase shift Δp is given by an Equation (18). Δ ⁢ ⁢ p = Arg ⁡ [ ∑ m = 0 M - 1 ⁢ I m ⁢ ⁢ exp ⁡ ( - 2 ⁢ ⁢ π ⁢ ⁢ ⅈ ⁢ ⁢ m M ) ] ( 18 ) where a symbol Arg represents a calculation of argument. A generated phase shift amount by the phase shifter 23 can be selected either an unequal interval or an equal interval. Phase-shift can be generated by taking a wedge shaped acrylic plate as shown in FIG. 9 in and out to the optical path by the phase shifter positioning mechanism 24, or by rotating a flat shaped acrylic plate in the optical path by the phase shifter positioning mechanism 24. These operations are performed by the controller 26. At the processing portion 27 shown in FIG. 6, (6) The combined image of the change in the amplitude and the phase is calculated (step 705—calculation of (ΔA+Δp) image) from the images obtained by the above described steps, and held in the processing portion 27. The composition ratio of the changes in the amplitude and the phase is given by a ratio of the change in the amplitude ΔA and the maximum value ΔAmax of ΔA. Namely, a composite value g is calculated by an Equation (19). g = Δ ⁢ ⁢ A Δ ⁢ ⁢ A max ⁢ Δ ⁢ ⁢ p + ( 1 - Δ ⁢ ⁢ A A max ) · Δ ⁢ ⁢ A ( 19 ) As mentioned above, the optimum imaging method is depending upon the density of the sample. Thus, (7) the amount of the density change in the sample such as a magnitude (an integrated amount (ΣΔA) for entire image) etc., are calculated (step 706—calculation of ΣΔA). At the display portion 28, an image which has the highest sensitivity and has been capable of imaging accurately from the images (the image of the change in the amplitude ΔA, the image of the change in the phase Δp, and the composed image of the changes in the amplitude and the phase) held in the processing portion 27, is displayed automatically according to the value of this ΣΔA. To put it more concretely, ΣΔA and A0 are compared with each other at step 707, and if ΣΔA is larger than A0, the image of ΔA is displayed (step 708). At step 709, ΣΔA and A1 are compared with each other, if ΣΔA is larger than A1, the image of Δp is displayed (step 710). If ΣΔA is smaller than both A0 and A1, the image of (ΔA+Δp) is displayed (step 712). Determination at each step 707 and 709 can be omitted, and when the calculations of 704 and 705 are finished, all of the images are sequentially displayed and/or the user select images to be displayed. FIG. 11A through FIG. 11C are views showing example images of biological sample obtained by using the embodiment 1. FIG. 11A shows the Δp image obtained at step 704′ in FIG. 8. This example cannot accurately show the phase-shift caused by the sample because of an unwrapping processing error. FIG. 11B shows the ΔA image obtained at step 704 in FIG. 8. This example (obtained by the present invention) can observe the image accurately even for the sample which has been impossible to observe the image accurately for the phase contrast method. FIG. 11C shows the (ΔA+Δp) image obtained at step 705 in FIG. 8. By combining the changes in the amplitude and the phase, it becomes possible to visualize not only the bone but also a structure inside the soft tissue by the image of one sheet, and an internal structure can further precisely be observed. As evident from the above description, according to the embodiment 1, the image of the change in the amplitude, the change in the phase, and the composite value of the changes in the amplitude and the phase of the interference beam as the contrast can be obtained as well. Therefore, even a sample including an organ of large density change such as the bone or the lung and an organ of small density change such as biological soft tissue can be observed in a state of high sensitivity. In the embodiment 1, the change in the amplitude and the phase of the interference beam caused by the sample has been detected by using the fringe scanning method. To perform this method, three sheets of the interference images which are in phase relationships different from each other are required at least. Because of its long measurement time, therefore it is difficult to apply this method to the dynamical observations. Accordingly, the embodiment using the Fourier transformation method in place of the fringe scanning method will be shown in this case. In the Fourier transformation method, the change in the amplitude and the phase can be obtained by one sheet of the interference image, therefore the measuring time can significantly be shortened. On the other hand, the spatial resolution is slightly lowered as compared with that of the fringe scanning method, because the spatial resolution is determined mainly by an interval of Moire-image interference fringes as described later. The same X-ray imaging apparatus shown in FIG. 6 can be adopted for the observation using the Fourier transformation method. To perform the Fourier transformation method, the wedge shaped phase shifter 23 shown in FIG. 9 which made of material having small absorption for X-ray such as the acrylic plate, is placed in one optical path of the interfering beams. As a result, the Moire interference fringes are formed in the interference image in a direction perpendicular to an inclined direction of the phase shifter 23. The intensity distribution of the interference image can be expressed as an Equation (20).I(x,y)=α(x,y)+c(x,y)exp(2πif0x)+c(x,y)exp(−2πif0x) (20) where c(c, y) is expressed by an Equation (21), c ⁡ ( x , y ) = 1 2 ⁢ A ⁡ ( x , y ) ⁢ ⁢ exp ⁡ ( ⅈ ⁢ ⁢ Δ ⁢ ⁢ p ⁡ ( x , y ) ) ( 21 ) where α is a background intensity distribution having no relation with an interference fringe, A is an amplitude of an interference fringe, Δp is a phase shift caused by the sample, f0 is a spatial frequency in an x direction of Moire interference fringes. Additionally, * indicates a complex conjugate. By the Fourier transformation calculation of Equation (20), frequency spectrum IF(x, y) in an x direction is given as an Equation (22).IF(f,y)=αF(f,y)+cF(f−f0,y)+cF(f+f0,y) (22) By setting the angle of the phase shifter 23 that the spacing of the Moire interference fringes is sufficiently narrowed as compared with a structure of the sample, αF, CF and CF become completely separated spectra in the Equation (22). Therefore, by performing the inverse Fourier transformation of the CF or CF* component after shifted by f0 in a direction of the origin, only c including information of the change in the amplitude A and the phase of the interference fringe can be obtained. Therefore, the amplitude A can be obtained by a calculation of an absolute value of c, and the phase Δp can be obtained by a calculation of the angle of deviation. The measurement can be performed by a procedure which is similar to that of the embodiment 1 (FIG. 8). The Fourier transformation method is used in place of the fringe scanning method for the measurement of the background data (step 701) and the actual measurement (step 703). The magnitude (integrated amount (ΣΔA) of the phase of the amplitude for entire image) etc. is also calculated in the same way as the embodiment 1. And according to this value, an image having best sensitivity and having been capable of imaging accurately is selected from the image of the measured change in the amplitude ΔA, the image of the measured change in the phase Δp, and the composed image of the change in the amplitude and the phase, and is displayed at the display unit 28. Or, all of the images are displayed so as to be selectable. As evident from the above description, according to the embodiment 2, the image of the change in the amplitude, the change in the phase, and the composite value of the changes in the amplitude and the phase of the interference beam as the contrast can be obtained in a short time. Therefore, even a sample including a organ of large density change such as the bone or the lung and a organ of small density change such as biological soft tissue can be observed in a state of high sensitivity and high time resolution. In the embodiments 1 and 2, only the image (transmitted image) passing through the sample can be measured. An embodiment capable of observing the inside of the sample nondestructively will be shown. FIG. 12 is a block diagram showing a constitution of the embodiment 3. The apparatus is similar to that of the embodiments 1 and 2 excluding a sample holder 33 and a sample holder rotating mechanism 34. The sample is fixed to the sample holder 33, and rotated in a direction perpendicular to an optical axis by the sample holder rotating mechanism 34. In order to reduce an influence by the shape of the sample, the inside of the sample holder 33 may be filled up with a liquid having the density close to the density of the sample. In the embodiment 3, the measurement is performed by the procedure shown in FIG. 13. (1) Similar to the procedure of the first half of FIG. 7, the step 1201 (measurement of background data), the step 1202 (placing the sample in the optical path), and the step 1203 (actual measurement) is executed, and at the step 1204, the change in the amplitude ΔA caused by the sample is obtained. (2) At step 1205, the sample is rotated by Δr by the sample rotating mechanism 34. (3) At step 1206, it is determined whether the sample has finished the rotation at the predetermined number of rotation. (4) When the sample has not finished the predetermined number of rotation, the sample is evacuated from the optical path at step 1207, then proceeds to step 1201 (measurement of background data), the measurement is repeated according to the similar procedures. (5) The above procedures (1) through (4) are repeated the required number of steps n (=180°/Δr), and when above iteration reached the predetermined number of rotation, the measurement is terminated. After the measurement, a cross sectional image of the change in the amplitude ΔA as the contrast is reconstituted (step 1204 (calculation of ΔA image)) from the measured data set. Similar to the embodiment 1, not only the change in the amplitude ΔA, but also the change in the phase Δp can simultaneously be measured at step 1203 (actual measurement). Thus, by utilizing these data, the image of the change in the amplitude, the change in the phase, and the composite value of the change in the amplitude and the phase can be obtained (step 1204′ (calculation of Δp image) and step 1204″ (calculation of (ΔA+Δp)image)) as shown in FIG. 14. Meanwhile, the composition of the amplitude A and Δp is carried out by a method similar to that of the embodiment 1. The magnitude (integrated amount (ΣΔA) of the phase of the amplitude for entire image) etc. is calculated in the same way as the embodiment 1. And according to this value, an image having best sensitivity and having been capable of imaging accurately is selected from the image of the measured change in the amplitude ΔA, the image of the measured change in the phase Δp, and the composed image of the change in the amplitude and the phase, and is displayed at the display unit 28. Or, all of the images are displayed so as to be selectable. As evident from the above description, according to the third embodiment 3, there can be provided the cross sectional image of the change in the amplitude, the change in the phase, and the composite value of the change in the amplitude and the phase as the contrast. Therefore, even a sample including an organ of large density change such as the bone or the lung and a organ of small density change such as biological soft tissue can be observed its internal structure nondestructively in a state of high sensitivity. Since the X-ray interferometer used in the embodiment 1 through 3 is constituted by one crystal block, the size of the interferometer is limited by a diameter of a crystal ingot. The field of view, therefore, could not widen 2 cm or more. As shown in FIG. 15, an example of the imaging apparatus having 2 cm or more field of view using two-crystal X-ray interferometer is given. In the embodiment 4, in order to widen the field of view, a two-crystal X-ray interferometer constituted by a first crystal block 38 and a second crystal block 39 having two wafers as shown in FIG. 15 is used. An incident X-ray 40 is split into a first beam 42 and a second beam 43 by X-ray diffraction of a Laue case by a first wafer 41 of the first crystal block 38. The first beam 42 is diffracted by a second wafer 44 of the first crystal block 38 and the second beam 43 is diffracted by a third wafer 45 of the second crystal block 39. Both diffracted beams incident at the same point on a fourth wafer 46 of the second crystal block 39, and combined to form a first interference beam 47a and a second interference beam 47b. The relative rotational fluctuation around Z axis between the first crystal block 38 and the second crystal block 39 causes the fluctuation of phase of the interference beam. Therefore, in order to perform stable imagines, above rotational fluctuation must be controlled with significantly good precision. A relationship between the fluctuation of mechanical rotation Δθ and the fluctuation of phase of interference beam Δφ is given by an Equation (23). Δ ⁢ ⁢ ϕ = 2 ⁢ ⁢ π ⁢ ⁢ Δ ⁢ ⁢ θ ⁡ ( x + t ) d ( 23 ) where t is a thickness of wafer of interferometer, x is an space between crystal wafers (between 41 and 44, 45 and 46), d is an Bragg-plane spacing, θB is a Bragg angle. For example, by utilizing diffraction of Si (220) (d=0.192 nm) under the condition of λ=0.07 nm, L=20 m, x=63 nm and θB=10.5°, it is calculated from Equation (23) that Δθ=2 nrad corresponds Δφ=2π approximately. Accordingly, in order to execute stable measurement, Δθ must be controlled at least by the precision of sub-n rad. In order to realize the above described positioning precision, positioning between the crystal blocks is carried out by a stage group (constituted by a first θ table 48 used for the θ axis rotation of entire interferometer, a second θ table 49 used for the θ axis rotation of the second crystal 39 and a tilt table 50 used for the rotation of a φ axis of the first crystal 38) using a sleeve bearing mechanism with slipping sheets to enhance the mechanical rigidity in the embodiment 4. A positioning mechanism 52 having the extraordinary high precision using a piezoelectric device etc. is used for a second θ stage 49 which is required for the sub-nrad positioning. In addition, a feedback positioning mechanism controlling the rotation angle between the crystal blocks by a feedback loop is employed for the suppressing of a drift of Δθ for a long time. The feedback mechanism uses a detector 51 for detecting the intensity of the first interference beam 47a, and when the intensity of the first interference beam is fluctuated due to drift rotation of Δθ, the feedback mechanism adjusts the rotation of the second θ table 49 via the positioning mechanism 52 so as to cancel the fluctuation at once. Furthermore, in the case of that above described control mechanism cannot sufficiently suppress the drift of Δθ resulting from the fluctuation of the intensity of the incident X-ray, a two dimensional X-ray detector is suitable for the detector 51. The feedback control is performed by adjusting the rotation of the second θ table 49 so as to cancel the motion of position of the Moire interference fringes. Each of a sample holder 53 and a sample holder positioning mechanism 54 has a function for rotating a sample similar to that of the embodiment 3. As a result, it is possible to observe the inside of the sample nondestructively by using the measurement similar to the embodiment 3. The measurement is carried out similar to the embodiments 1 and 3. The magnitude (an integrated amount (ΣΔA) of the phase of the amplitude for entire image) etc. is calculated. According to this value, an image having best sensitivity and having been capable of imaging accurately is extracted from the images (the image of the change in the amplitude ΔA, the image of the change in the phase Δp, and the composed image of the change in the amplitude and the phase) held at the processing portion 27 and is displayed at the display unit 28. As evident from the above description, according to the fourth embodiment 4, even the size of a sample is more than 2 cm, there can be obtained the transmission image and the cross sectional image of the change in the amplitude, the change in the phase, and the composite value of the change in the amplitude and the phase of the interference beam as the contrast. Even a sample in which a portion having a large change in density and a portion having a small change in density are mixed, the sample can be observed in a state of high sensitivity. According to the present invention, the imaging apparatus and imaging method using the amount of change in the amplitude caused by the sample as the contrast and/or the amount composed of the amount of change in the amplitude and the amount of phase-shift caused by the sample as the contrast, enables to observe a sample including a organ of large density change such as the bone or the lung and a organ of small density change such as biological soft tissue, which has heretofore been difficult by means of the conventional adsorption and the phase contrast X-ray imaging apparatuses, in a state of high-sensitivity. Reference numerals in drawings are as follows: 1: beam splitter, 2: mirror, 3: analyzer, 4: X-ray, 5 and 6: beam, 7 and 8: interference beams, 9: sample, 10: X-ray interferometer, 11: splitter, 12: mirror optical system, 13: analyzer, 14: incident X-ray beam, 15: first beam, 16: second beam, 17: interference beam, 18: sample, 19: X-ray interferometer, 20: position adjusting mechanism, 21: sample holder, 22: sample holder positioning mechanism, 23: phase shifter, 24: phase shifter positioning mechanism, 25: X-ray detector, 26: controller, 27: processing portion, 28: display unit, 29: incident X-ray, 30: splitter, 31: mirror, 32: analyzer, 32a: first interference beam, 32b: second interference beam, 33: sample holder, 34: sample holder rotating mechanism, 38: first crystal block, 39: second crystal block, 40: incident X-ray, 41: fist wafer, 42: first beam, 43: second beam, 44: second wafer, 45: third wafer, 46: fourth wafer, 47a: first interference beam, 47b: second interference beam, 48: first θ table, 49: second θ table, 50: tilt table, 51: detector, 52: positioning mechanism, 53: sample holder, 54: sample holder positioning mechanism.
	summary
043127746	description	The following examples are presented. Unless otherwise specified all solutions are aqueous solutions. The "aqueous ammonium hydroxide" or "NH.sub.4 OH" used in the Examples contained about 28% NH.sub.3, ppm means parts per million parts of solution, ppb means parts per billion parts of solution, ppt means parts per trillion parts of solution, all parts and percentages are on a weight basis and all temperatures are given in degrees Centigrade. For reasons of safety all simulated radwaste solutions used in the Examples were actually non-radioactive; however, radioactive solutions of the same kind can be substituted and concentrated and encapsulated in accordance with the following Examples. EXAMPLE 1 Preparation of Glass Particles and Tubes A. A molten glass was formed in a platinum crucible at 1400.degree. C. from sand, boric acid, sodium carbonate and potassium carbonate, the glass having a nominal composition of 3.5 mole percent Na.sub.2 O, 3.5 mole percent K.sub.2 O, 33 mole percent B.sub.2 O.sub.3 and 60 mole percent SiO.sub.2. The molten glass was vertically updrawn and solidified into rods having a diameter of about 0.8 cm and a length of about 100 cm which were then crushed in a stainless steel cylinder with a stainless steel rod. The resulting powder was sieved and the fraction between 32 and 150 mesh screens was selected for use in certain of the following Examples. B. Tubes were formed by pulling the above-described molten glass and applying a small internal pressure. Tubes that were sealed at one end were formed by turning off the internal pressure during the drawing operation. Tubes open at both ends were formed by maintaining the internal pressure through the drawing and cut-off operation. The tubes were formed with an outside diameter of about 1 cm and a wall thickness of about 0.15 cm ad were cut to about 5 cm long. EXAMPLE 2 Preparation of Porous Glass Tubes A base glass tube having one sealed end and one open end was prepared as described in Example 1B. The tube was then heat-treated at 550.degree. C. for 110 minutes in an electric furnace to induce suitable phase separation. The tube after heat-treatment was annealed by cooling slowly down to room temperature, and was leached to form a porous tube by soaking it in a 3 N HCl solution saturated with NH.sub.4 Cl at 95.degree. C. for two days. The porous tube was then soaked in hot water for one day to wash out residue from the leaching operation and was then kept in a dessicator until the pores were dry of the washing water. The resulting porous glass tube had a nominal composition of 95 mole percent SiO.sub.2, 5 mole percent B.sub.2 O.sub.3 having interconnected pores, and an internal surface of about 100 m.sup.2 /gr. The surface of the resulting porous glass tube was saturated with .tbd.SiOH groups. EXAMPLE 3 Preparation of Porous Glass Powder Glass rods were prepared as described in Example 1A. Before crushing the glass rods, they were heat-treated at 550.degree. C. for 110 minutes and then crushed to form glass powder. Next the glass powder was sieved and the fraction passing through a 32 mesh screen but not through a 150 mesh screen was leached in a 3 N HCl solution at about 95.degree. C. for about six hours. The glass powder was washed with deionized water for about 24 hours at about 25.degree. C. The resulting porous glass powder had a nominal composition of 95 mole percent SiO.sub.2 ; 5 mole percent B.sub.2 O.sub.3, had interconnected pores, and had an internal surface of about 100 m.sup.2 /gr. The resulting glass surface was saturated with SiO H groups. The porous glass powder was dried in a beaker on a hot plate at about 150.degree. C. EXAMPLE 4 Use of a Porous Glass Tube to Concentrate and Encapsulate A dry porous tube having one open end and one closed end, prepared as described in Example 2, was impregnated with a solution containing dissolved CsNO.sub.3 and Al.sub.2 O.sub.3 particles simulating a nuclear waste fluid. The CsNO.sub.3 solution contained 67 grs of CsNO.sub.3 (which could be radioactive) dissolved in 23 ml water at 100.degree. C. and 10 grs of Al.sub.2 O.sub.3 representing suspended solids (which could be contaminated with radioactive isotopes). The interior of the tube was filled with the dopant solution, and the solution was allowed to penetrate into the pores. Some of the solution in the tube was allowed to pass through the tube walls to the outside of the tube and was collected for use in other tubes. This was continued until the interior of the tube was essentially empty of the solution. The Al.sub.2 O.sub.3 solids suspended in the solution, however, being much larger than the pore size of the tube walls were retained in the interior of the tube. Also, the solution containing the dissolved CsNO.sub.3 filled the pores of the glass tube walls. The resulting laden porous tube was then inserted in methanol at 0.degree. C. to cause the dissolved CsNO.sub.3 in the solution in the pores to precipitate in the pores. The inner and outer surfaces of the laden tube were soaked in clean methanol at 0.degree. C. for 24 hours, while changing the methanol often, resulting in thin layers on both the outside and inside surfaces of the tube in which the concentration of the precipitated CsNO.sub.3 was lower than the concentrations of precipitated CsNO.sub.2 deeper in the glass. (That is the inner and outer surface layers or regions contained approximately one fifteenth of the CsNO.sub.3 concentration of regions located deeper in the tube wall). The porous tube was then removed from the 0.degree. C. methanol bath and placed into a larger diameter (3.5 cm), substantially non-porous, fused silica glass tube having an open end and was dried under vacuum at 0.degree. C. for 24 hours. The fused silica glass tube containing the laden porous tube was then allowed to warm under vacuum to room temperature and was put into a furnace where it was slowly heated at 15.degree. C./hr up to 625.degree. C. This heating period allowed the pores of the glass to dry further. The laden porous tube inside the non-porous tube was held at 625.degree. C. for 16 hours to ensure that all the CsNO.sub.3 was decomposed and the resulting nitrogen oxides were expelled leaving Cs.sub.2 O. It was then heated to 875.degree. C. still under vacuum in order to fuse the pores and sinter the glass structure of the porous glass tube thus converting it into a substantially non-porous glass tube with the cesium (Cs.sub.2 O) trapped as a part of the glass structure. The solid (Al.sub.2 O.sub.3) remained deposited on the tube interior. The tube is placed horizontally on a graphite block in a ceramic tube furnace with another graphite block resting on top of it. It is heated to about 1350.degree. C. and the tube sags under the weight of the upper graphite block causing the interior surfaces of the tube to fuse and seal together, thus immobilizing and encapsulating both the Cs.sub.2 O from originally dissolved CsNO.sub.3 and the originally dispersed Al.sub.2 O.sub.3 solids. EXAMPLE 5 Use of Porous Powder in Non-Porous Tube to Encapsulate A non-radioactive aqueous solution simulating a radwaste stream projected for an existing spent nuclear fuel reprocessing plant and containing 3.06 grs Fe(NO.sub.3).sub.3.9H.sub.2 O, 1.68 grs Ce(NO.sub.3).sub.3.6H.sub.2 O, 0.78 grs La(NO.sub.3).sub.3.6H.sub.2 O, 0.78 grs CsNO.sub.3, 3.88 grs Nd(NO.sub.3).sub.3.5H.sub.2 O, 0.52 grs Ba(NO.sub.3).sub.2, 2.72 grs Zr(NO.sub.3).sub.4, 0.42 grs Sr(NO.sub.3).sub.2, 0.34 grs Y(NO.sub.3 0.sub.3.5H.sub.2 O and 5 ml water, with all elements in solution except Zr(NO.sub.3).sub.4 which was present as a precipitate, was poured into a 50 ml beaker which contained 5 grs of porous glass powder made as described in Example 3. The excess solution was decanted and the beaker was heated to 200.degree. C. on a hot plate to dry the glass powder and deposit the dissolved nitrates in the pores of the glass powder and the undissolved Zr(NO.sub.3).sub.4 on the outer surfaces of the glass powder. The laden glass powder was then placed in a Vycor tube (Corning 743170-4381) having a nominal composition of 96% SiO.sub.2 and 4% B.sub.2 O.sub.3, an inside diameter of 7 mm, an outside diameter of 9 mm and a length of 50 cm. The tube was sealed at one end and was connected to a vacuum pump. The tube containing the laden porous glass powder was then inserted into a furnace at room temperature under vacuum and heated at 15.degree. C./hr up to 600.degree. C. to evaporate any remaining water or other volatiles and to decompose the nitrates present into the corresponding metal oxide and nitrogen oxides and to expel the nitrogen oxides. After holding at 600.degree. C. for 24 hours., the tube was transferred to a second furnace capable of providing higher temperatures. Upon transferring from one furnace to the other, the temperature dropped to 530.degree. C. The temperature in the second furnace was increased gradually from 530.degree. C. to 1340.degree. C. over a period of three hours and 25 minutes. The tube was removed and was found to have collapsed above the level of the glass powder which had been impermeated with the simulated nuclear waste solution. This occurred because the furnace had a relatively large temperature gradient across it, and the tube had been inserted too far. Nevertheless, the final product was a partially collapsed tube completely sealing within it the glass powder with no cracks present in the tube. The uncollapsed lower portions of the tube contained the impermeated glass some of which was a loose powder, some of which had melted into chunks and some of which had melted and stuck to the interior walls of the tube. There were no breaks in the tube walls and no stress of the tube walls was observed under crossed polaroids. The resulting product effectively encapsulated the metal oxides resulting from the metal nitrates in the initial simulated nuclear waste stream and isolated them from the environment. EXAMPLE 6 Use of a Porous Glass Powder In a Non-Porous Glass Tube to Encapsulate Porous glass powder prepared in the manner described in Example 3 was poured into a 100 ml beaker and soaked at room temperature for 17 hours in a basic solution of 12 ml NH.sub.4 OH and 14 gr NaNO.sub.3 in 38 ml of water. Ion exchange took place as described in the above-mentioned concurrently filed application to form silicon-bonded sodium oxy groups on the outer surfaces of the glass powder particles and in the pores of the glass powder. The solution was then poured off and the porous glass powder was rinsed to a neutral pH. A simulated radwaste solution containing 12 grs Cu(NO.sub.3).sub.2 and 12 grs CsNO.sub.3 in 73 ml of water was poured into the beaker to cover the glass powder and the latter was allowed to soak for 16 hours. The solution was then removed leaving behind Cs and Cu ions bonded to the glass through silicon-bonded oxy linkages. The laden powder was rinsed in water and dried on a hot plate at about 200.degree. C. for one hour. It was then poured into a Vycor tube of the kind described in Example 5. The tube and contents were evacuated in vacuum and further dried in a furnace at room temperature and then heated to 450.degree. C. overnight in an electric furnace. It was then heated according to the schedule given in Table 1 below: TABLE 1 ______________________________________ Time (Hour:Minute) Temperature, .degree.C. Pressure, m Torrs ______________________________________ 9:12 450 4 9:25 680 -- 9:42 1230 4 9:50 1350 4 ______________________________________ Between 9:25 and 9:42, the porous glass beads sintered and became non-porous trapping both Cu ans Cs ions within the resulting glass structure. The glass tube was further heated while a mechanical vacuum pump was holding the pressure to 4 m torrs and at about 1350.degree. C. the tube collapsed. The tube was withdrawn from the furnace, and cooled. After sufficient time, both the outside Vycor glass and the inside sintered glass powder cooled below their respective glass transition temperatures and stress built up between both glasses. On further cooling the Vycor cracked. In order to prevent cracking, the original outside glass tube should have a higher thermal expansion coefficient to match or approximate the coefficient of the interior mass of sintered glass powder and Cu and Cs ions in it. Nevertheless, the tube did perform the important function of containing the potentially radioactive vapors, e.g., CsNO.sub.3, Cu, CuNO.sub.3 vapors, during processing. These vapors condensed on the interor upper surfaces of the tube and thus were prevented from entering the environment. EXAMPLE 7 Use of Cation Exchange Porous Glass Powder to Concentrate Dilute Radwaste and Encapsulate Radioactive Materials Removed Therefrom When handling low level nuclear waste, the concentrations of the dissolved and undissolved radioactive materials in the solution can be very low. Additives to prevent corrosion, often make such waste streams basic. A simulation of such a stream was performed by adding 0.011 g CsNO.sub.3, 0.018 g Cu(NO.sub.3).sub.2, 12.5 ml aqueous NH.sub.4 OH solution (28% NH.sub.3) to water to form 50 ml of solution. The porous glass powder prepared as described in Example 3 was added to the above solution and stirred for three hours. The cation exchange porous glass powder became pale blue due to the exchange of protons (of the hydroxyl groups on the internal and external surfaces of the porous glass powder) by copper cations which then became bonded to silicon through oxy linkages. The solution was removed from the glass powder which was then rinsed in water and dried on a hot plate at about 200.degree. C. for one hour. The powder was poured into a Vycor tube of the kind described in Example 5 and the tube and contents were evacuated under vacuum and gradually heated in a furnace from room temperature (about 20.degree.C.) to about 1350.degree. C. over a period of about three hours and 5 minutes. During heating the pores of the glass powder collapsed and the powder sintered into an integral mass trapping the Cs and Cu ions within it. During the last stages of heating the tube collapsed. Because the amount of simulated nuclear waste (Cs and Cu) encapsulated in the porous glass powder was small, the viscoelastic properties of the tube and the sintered glass powder matched each other more closely than in Example 6, and the final glass capsule did not crack. It remains monolithic and provided a continuous cladding around the mass of sintered glass powder containing the simulated radioactive Cs and Cu. EXAMPLE 8 Encapsulation of Calcined Nuclear Waste in a Vycor Tube for Burial About 1.5 ml of a non-radiactive aqueous solution simulating a radwaste stream projected for a spent nuclear fuel reprocessing plant and as described in Example 5 were placed in a 50 cm long Vycor tube which also is described in Example 5. The solution included dissolved nitrates as well as precipitated Zr(NO.sub.3).sub.4 as described in Example 5. No glass powder was added. The tube was connected to a vacuum pump by a rubber hose. In order not to have excessive bubbling, the tube was placed in an ice bath at 0.degree. C. and pumped overnight to dry its contents. The next day the temperature of the tube was 28.degree. C. and the interior pressure was 20 m Torrs. The tube was transferred to a furnace where it was heated under vacuum according to the heating schedule given in Table 2 below. TABLE 2 ______________________________________ Time (Hours:Minute) Temperature, .degree.C. Pressure, m Torr ______________________________________ 12:45 70 137 13:40 80 40 13:50 130 140 14:05 155 50 14:25 190 79 14:50 190 25 15:15 290 50 15:30 340 80 15:40 350 55 16:05 450 34 17:05 600 16 18:10 850 16 20:00 1340 14 ______________________________________ At 20:00, after seven hours and 15 minutes of heating, the tube which had collapsed during heating, was removed from the furnace. From the data in the above Table, it can be seen that pressure maxima occurred at 12:45, 13:50 and 14:25. This appears to have been due to the evaporation of water still in the tube when it was placed in the furnace and appears to have occurred each time when the temperature was significantly raised. If the temperature is held constant as at 13:40, 14:05 and 14:50, the pressure is reduced as the water vapor is taken off by the vacuum. Another maximum occurs around 15:30 at about 300.degree.-400.degree. C. which is apparently due to the decomposition of nitrates to form nitrogen oxides. The final product was a collapsed and sealed Vycor tube with calcined simulated nuclear waste (i.e., the oxides Fe, Ce, Ha, Cs, Nd, Ba, Zr, Sr and Y) encapsulated inside the collapsed and sealed tube. The surface of the collapsed and sealed tube showed no cracks. When the tube was examined under polarized light it was found to be free of stress. The resulting product was suitable for burial in the ground or sea and can be packaged with other like products in larger containers for such purposes. EXAMPLE 9 Use of Non-Porous Glass Powder in a Non-Porous Glass Tube For Incapsulating Nuclear Waste For Burial Pyrex glass (Corning 234030-510) having a nominal composition of 81%, SiO.sub.2, 2% Al.sub.2 O.sub.3, 13% B.sub.2 O.sub.3 and 4% Na.sub.2 O (given in wt. %'s) was crushed in a stainless steel cylinder using a stainless steel rod. The crushed glass was sieved and the fraction which passed through 60 mesh and was caught on 150 mesh was selected for use. 9.5 Gms of the selected fraction of Pyrex powder were mixed with 0.5 gm of porous glass powder impregnated with simulated nuclear waste stream and dried as described in Example 5. The mixed powder was further dried in a beaker on a hot plate at 110.degree. C. for about two hours. Part of this mixed powder was then placed in a 50 cm long Pyrex tube having the nominal composition given above, a 9 mm O.D. and a 7 mm I.D., so that it formed a column 10 cm high. Also, a piece of platinum wire, 1 cm long and 1.5 mm in diameter was added to the powder in the tube. The open end of the tube was attached to a vacuum pump and placed in a furnace where it was gradually heated from about 25.degree. C. to about 830.degree. C. in about four hours and 35 minutes. The finished product developed some cracks after it was pulled out of the furnace. The cracks appeared to be internal and did not extend to the outside surface of the collapsed Pyrex tube. The resulting product effectively encapsulated the glass powder containing simulated radioactive waste materials and platinum which represented the platinum group metals such as Pd, Ru and Rh that are commonly dispersed solids in nuclear waste streams. The cracking can be eliminated by more closely matching the thermal expansion coefficient of the tube and of the contents. The final product can be suitably buried underground or at sea, preferably with other like products and packaged in a larger container for convenience. EXAMPLE 10 Trapping Radioactive Vapors In A Porous Glass Rod The purpose of this example is to show that gas products emanating from the simulated nuclear waste being heated in a glass tube can be trapped in a porous glass rod. 6 Gms of porous glass powder prepared as described in Example 3, was mixed in a beaker with 2.76 gms of CsNO.sub.3, 3.17 gms of Cu(NO.sub.3).sub.2, 73 ml of H.sub.2 O and 25 ml of NH.sub.4 OH for 20.5 hours and washed for 24 hours. The impregnated porous glass powder was dried on a hot plate at a low temperature (about 200.degree. C. for about one hour). Then, the sample was placed in a Pyrex glass tube identical to the one described in Example 9 and having one end closed and a constricted neck located about 11 cm from the closed end. The powder formed a 4 cm high column in the tube. A 12.5 cm long porous glass rod, as prepared in Example 1A, having a diameter slightly less than 7 mm was inserted into the tube. The inner end of the rod had been ground down to a taper shape (which then was washed in a HF solution to free the pores) so that a fairly good seal was made between this end of the rod and the constricted neck section of the tube. The tube was placed upright partly inside a furnace so that the upper half of the rod was outside the furnace. Heating was carried out according to the time, temperature and pressure schedule shown in Table 3 below. At the end of the heating cycle, the tube was removed from the furnace. The bottom portions of the tube had collapsed up to 1 cm below the tapered end of the porous rod. The 5 cm section of the rod which was half inside and half outside the furnace was slighty yellow in color indicating the condensation of copper vapors, while all the other parts of the tube and the rod were substantially colorless. This indicates that the simulated radioactive vapor, i.e., copper vapors, escaping from the impregnating porous glass powder during tht heating process were trapped in the approximately 5 cm section of the porous rod and prevented from leaving the tube. The resulting collapsed tube product effectively encased the simulated radwaste in a strong glass structure. TABLE 3 ______________________________________ Time (Hours:Minute) Temperature, .degree.C. Pressure, m Torrs ______________________________________ 2:30 20 5 2:31 95 22 2:52 95 17 3:13 95 13 3:43 150 13 4:21 260 24 9:30 580 8 10:20 750 12 ______________________________________ The pressure maxima at 2:31 is due to water being expelled from the porous glass powder and the maxima at 4:21 is due to the nitrogen oxides produced by the decomposition of the cesium and copper nitrates. EXAMPLE 11 Porous Rod in Non-Porous Tube A porous rod having a length of 2.5 cm and a diameter of 0.7 cm and made in a manner similar to that described in Example 1A but not crushed and phase-separated and acid-leached as described in Example 2 was dried under vacuum, at room temperature, for 48 hours. The rod was then immersed (for 24 hours) in a 90.degree. C. solution comprising 15.28 g Fe(NO.sub.3).sub.3.9H.sub.2 O+2.16 g Sr(NO.sub.3).sub.2 +2.03 g (Y(NO.sub.3).sub.3.6H.sub.2 0+13.61 g Zr(NO.sub.3).sub.4 +3.98 g Cs(NO.sub.3)+2.67 g Ba(NO.sub.3).sub.2 +3.93 g La(NO.sub.3).sub.3.6H.sub.2 O+8.38 g Ce(NO.sub.3).sub.3.6H.sub.2 O+19.45 g Nd(NO.sub.3).sub.3.5H.sub.2 O67 ml H.sub.2 O. This solution simulates a projected waste stream of a nuclear reactor power plant. During the immersion, the solution diffused into the porous glass rod and filled its pores. The rod was then immediately transferred into a Pyrex test tube (nominal composition given in Example 9, 1.4 cm in O.D. and 15 cm long) containing a 110.degree. C. solution whose composition is similar to the one mentioned above. Vacuum was then applied to the tube in order to speed up the drying of the rod by evaporating the water which was present in the solutions. After a few minutes, the nitrates were observed to have precipitated inside the rod. A small amount of the nitrates precipitated on the surface of the rod forming a thin film which was later mechanically removed. To insure the completeness of the drying, the rod was left under vacuum at 110.degree. C. for about three hours. The dried impregnated rod was then placed under vacuum in a Vycor tube (having the nominal composition given in Example 5, 1 cm in O.D. and 50 cm long), at room temperature. The tube was then heated from room temperature to 625.degree. C. at 15.degree. C./hr., and from 625.degree. C. to 890.degree. C. at 50.degree. C./hr. (at which latter temperature the pores of the rod collapsed). The rod and first Vycor tube were then placed in a second Vycor tube (having the nominal composition given in Example 5, 50 cm long with a 15 mm O.D. and 12 mm I.D.) which was closed at one end. Then, the rod and tubes were placed in a furnace and heated from450.degree. C. to 1340.degree. C. over a period of two hours and 26 minutes. After this time, the tube and rod assembly were taken out of the furnace. The section of the tube around the stuffed rod had collapsed on the rod as well as an approximate 1.5" section above the rod. /The rod seemed to have some cracks in the rod which actually were there prior to heating, while the tube did not appear to have any cracks. EXAMPLES 12-19 Each of the Examples 4 through 11 are repeated, except that corresponding radioactive nitrates are used in place of the corresponding non-radioactive nitrates specified in Examples 4 through 11 and radioactive by contaminated Al.sub.2 O.sub.3 is used in place of non-radioactive Al.sub.2 O.sub.3 specified in Example 4. In each instance, the radioactive material is immobilized and encapsulated within the resulting glass product. EXAMPLES 20 and 21 In Example 20, silica gel purchased from DuPont as Ludox HS-40% is poured into a Vycor glass* tube plugged with glass wool at the top opening and a porous glass disc at the bottom opening to prevent the silica gel particles from escaping. Analysis of the silica gel (Ludox HS-40%) by atomic absorption before starting shows that it contains 40 wt.% SiO.sub.2, 0.41 wt. % titratable alkali as Na.sub.2 O, less than 0.1 wt. % Cs.sub.2 O, and a ratio of SiO.sub.2 to Na.sub.2 O of about 95 to 1. The silica gel contains about 0.4 mole percent silicon-bonded sodium oxy groups. The titratable sodium content is believed to be in the form of silicon-bonded surface sodium oxy groups and the surface to weight ratio is about 230 m.sup.2 /g. An essentially neutral solution containing 10 g CsNO.sub.3 per 100 ml of water is passed slowly through the silica gel. After several liters are passed through, the silica gel is dried and heated in vacuum above 100.degree. C. until the silica gel is observed to sinter (below 1000.degree. C.) and then it is heated further with a vacuum in the Vycor tube to collapse the tube on the sintered silica gel (which normally occurs below 1300.degree. C.). The final product, after cooling to room temperature is a solid rod with the outside surface consisting of at least 94% silica (e.g., the composition of the Vycor tube), and an interior containing Cs bonded to silicon through oxygen linkage and fused into the structure. Cesium oxide content is analyzed by atomic absorption spectroscopy to be at or above 2 weight percent based on the weight of silica gel. Sometimes the rods break into several pieces, but the immobilization and containment of the radwaste in the resulting glass product is still many times better than what is obtained by the prior art methods using cement. Vycor brand silica glass No. 7913 made by Corning Glass Works and containing 96 wt. % silica and 4 wt. % B.sub.2 O.sub.3. Example 21 is carried out in exactly the same manner as described above with the only exception being that the CsNO.sub.3 is radioactive. There results a final product in which the radioactive Cs is chemically bonded through oxy linkages to silicon of the silica gel within the collapsed Vycor tube which effectively encapsulates and seals the radioactive Cs from the environment. EXAMPLES 22 and 23 Porous glass beads having an average diameter of about 50 to about 100 microns are prepared by the process described in Example 1A except that instead of pulling the molten glass into rods, it is quenched by pouring it into a cooling bath of water so as to form small fractured glass particles (frit) of varied shapes. The glass particles or frit are then formed into spheres by passing them through a radiant heating zone or high temperatue flame where they soften sufficiently to permit surface tension forces to form them into spheres while they are freely moving through the air. They are then cooled rapidly to prevent deformation or devitrification. There are thus formed beads or spheres averaging about 50 to about 100 microns in diameter. These beads or spheres are treated in the manner described in Example 3 to provide porous silicate glass beads and are subjected to a primary ion exchange treatment in a 3.2 M sodium nitrate-ammonium hydroxide solution for three days followed by rinsing well with deionized water until the pH of the rinse water is reduced to about 8. The beads so treated contain 2.0 wt. % silicon-bonded sodium oxy groups expressed as Na.sub.2 O, i.e., 4.0 mole percent sodium cations bonded to silicon through oxy linkages on the inner surfaces of the pores thereof. An ion exchange column of Vycor glass (Example 5) is plugged with a porous glass disc at the bottom and is filled with the treated spheres or beads, i.e., the beads having the silicon-bonded sodium oxy groups. A radioactive waste stream containing undissolved radioactive solids and dissolved radioactive Cs.sup.+ cations, in Example 27, or containing radioactive Sr.sup.2+ ions, in Example 23, is passed through the column. In each case, the aqueous solution coming from the bottom of the column is substantially free of radioactive cations. After a suitable period of time to provide an adequate loading of the radioactive material on and in the pores of the beads, the waste stream is diverted to a similar column. The loaded column is then heated by a heating zone traveling from the bottom up to first dry it, then to decompose the nitrates and drive off the nitrogen oxide decomposition products, then to close the pores of the beads, then to sinter the beads and finally to collapse the hollow Vycor column on the sintered beads to trap and encapsulate the sintered beads within the collapsed Vycor column. The radioactive cations are bound to the interior of the sintered bead mass and the undissolved radioactive solids are also trapped on the interior of the sintered bead mass as well as between the sintered bead mass and the interior collapsed Vycor column, thus providing a durable, leach-resistant glass product containing the radioactive waste materials and which is suitable for burial. EXAMPLE 24 This example illustraes a method for treating primary coolant from a pressurized water nuclear reactor plant. A mixture of powders of silica, boric acid, sodium carbonate and potassium carbonate is prepared in such proportions that yield a glass comprising 3.5 mole percent Na.sub.2 O, 3.5 mole percent K.sub.2 O, 33 mole percent B.sub.2 O.sub.3 and 60 mole percent SiO.sub.2. The mixture is heated in a platinum crucible up to 1400.degree. C. in an electric furnace to produce a molten glass which is pulled into rods about 8 mm in diameter and about 2.5 cm long. The glass rods are cooled and the glass is phase-separated by heat treating at about 550.degree. C. for about 110 minutes. The rods are then crushed to form a powder which is sieved through a 32 mesh screen onto a 150 mesh screen. The glass particles collected on the 150 mesh screen are leached in 3 N HCl at about 50.degree. C. for about 6 hours to remove the boron-rich phase and leave behind a porous glass comprising about 95 mole percent SiO.sub.2 and about 5 mole percent B.sub.2 O.sub.3. The porous glass has interconnected pores and contains at least about 5 mole percent silicon-bonded hydroxyl groups. The glass particles are then rinsed in deionized water until the rinse water reaches a pH of about 7. The porous glass powder is then immersed in an approximate 3.2 molar sodium nitrate-ammonium hydroxide aqueous solution for three days and then is rinsed in water until the pH of the rinse water is reduced to about 8. The resulting powder is then placed in an ion exchange column made of the Vycor glass as described in Example 5. A radioactive primary coolant from a pressurized water nuclear reactor plant utilizing UO.sub.2 fuel clad in stainless steel (containing 4.9 weight percent .sup.235 U) is passed through the column. The primary coolant has the composition given in Table 4 below which lists the radionuclide, the probable source, the probable form and the average concentration in microcuries per milliliter. The cationic radionuclides ion-exchange with sodium cations bonded to silicon through oxy groups in the porous silicate glass powder. TABLE 4 ______________________________________ Average Average Radio- Probable Probable Concentration Concentration nuclide Source.sup.a Form.sup.b (.mu.Ci/ml) (ppb) ______________________________________ 3.sub.H (1), (2) Water, gas 2.4 0.249 14.sub.C 1.2 .times. 10.sup.-5 2.69 .times. 10.sup.-3 24.sub.Na (1) Cation 1.9 .times. 10.sup.-2 2.18 .times. 10.sup.-6 32.sub.P 3.3 .times. 10.sup.-5 1.16 .times. 10.sup.-8 35.sub.S 3 .times. 10.sup.-6 7.08 .times. 10.sup.-8 51.sub.Cr (1) Anion 3.7 .times. 10.sup.-4 4.02 .times. 10.sup.-6 54.sub.Mn (1) Cation, s 2.7 .times. 10.sup.-4 3.38 .times. 10.sup.-5 55.sub.Fe (1) Cation, s 1.9 .times. 10.sup.-4 7.6 .times. 10.sup.-5 59.sub.Fe (1) Cation, s 1.0 .times. 10.sup.-5 2.03 .times. 10.sup.-7 57.sub.Co (1) Cation, s 1.2 .times. 10.sup.-6 1.42 .times. 10.sup.-7 58.sub.Co (1) Cation, s 4.7 .times. 10.sup.-4 1.48 .times. 10.sup.-5 60.sub.Co (1) Cation, s 7.7 .times. 10.sup.-5 6.81 .times. 10.sup.-5 63.sub.Ni (1) Cation, s 8.0 .times. 10.sup.-6 1.30 .times. 10.sup.-4 64.sub.Cu (1) Cation, anion, s 5.4 .times. 10.sup.-4 1.41 .times. 10.sup.-7 89.sub.Sr (2) Cation 2.8 .times. 10.sup.-6 9.93 .times. 10.sup.-8 90.sub.Sr (2) Cation 4 .times. 10.sup.-7 2.84 .times. 10.sup.-6 91.sub.Sr (2) Cation 9.8 .times. 10.sup.-5 2.76 .times. 10.sup.-8 90.sub.Y (2) s 91.sub.Y (2) s 92.sub.Y (2) s 95.sub.Zr (1), (2) s 1.7 .times. 10.sup.-5 8.06 .times. 10.sup.-7 95.sub.Nb (1), (2) s 1.9 .times. 10.sup.-5 4.83 .times. 10.sup.-7 99.sub.Mo (1), (2) Anion 1.2 .times. 10.sup.-4 2.54 .times. 10.sup.-7 103.sub.Ru (2) s 0 106.sub.Ru (2) s 0 122.sub.Sb (1) s 1.0 .times. 10.sup.-4 2.62 .times. 10.sup.-7 124.sub.Sb (1) s 2.0 .times. 10.sup.-5 1.16 .times. 10.sup.-6 132.sub.Te (2) Anion, s 131.sub.I (2) Anion 4.6 .times. 10.sup.-5 3.71 .times. 10.sup.-6 132.sub.I (2) Anion 133.sub.I (2) Anion 6.2 .times. 10.sup.-4 5.5 .times. 10.sup.-7 135.sub.I (2) Anion 9 .times. 10.sup.-4 2.60 .times. 10.sup.-7 134.sub.Cs (2) Cation 4.7 .times. 10.sup.-7 3.62 .times. 10.sup.-7 136.sub.Cs (2) Cation 0 137.sub.Cs (2) Cation 1.1 .times. 10.sup.-6 1.26 .times. 10.sup.-5 140.sub.Ba (2) Cation 4.7 .times. 10.sup.-6 6.45 .times. 10.sup.-8 141.sub.Ce (2) Anion, s 0 143.sub.Ce (2) Anion, s 0 144.sub.Ce (2) Anion, s 0 143.sub.Pr (2) Anion, s 110m.sub.Ag (1) s 1.2 .times. 10.sup.-5 2.52 .times. 10.sup.-6 181.sub.Hf (1) s 6 .times. 10.sup.-6 3.70 .times. 10.sup.-7 182.sub.Ta (1) s 2.5 .times. 10.sup.-5 4.01 .times. 10.sup.-6 183.sub.Ta (1) s 6.2 .times. 10.sup.-5 4.34 .times. 10.sup.-7 185.sub.W (1) s 1.2 .times. 10.sup.-5 1.28 .times. 10.sup.-6 187.sub.W (1) s 3.7 .times. 10.sup.-4 5.30 .times. 10.sup.-7 85m.sub.Kr (2) Gas 85.sub.Kr (2) Gas 88.sub.Kr (2) Gas 133.sub.Xe (2) Gas 8.9 .times. 10.sup.-5 4.78 .times. 10.sup.-8 135.sub.Xe (2) Gas 9 .times. 10.sup.-5 3.54 .times. 10.sup.-8 ______________________________________ .sup.a (1) Neutron activation products of nuclides from fuel cladding, construction material, and water. (2) Leakage from fuel. Mostly fission products. .sup.b Gas: presumably as dissolved gas. s: insoluble solids. The radioactive cations of the radionuclides listed in Table 4, cation-exchange with sodium cations bonded to silicon through oxy groups in the porous glass thereby binding the radionuclides to the porous glass through said silicon-bonded oxy groups and releasing non-radioactive sodium cations to the coolant solution. The insoluble radioactive solids in the coolant also filter out on the external surfaces of the porous glass particles. Additional porous glass particles can be added to increase the filtering capacity of the ion exchange column as the insoluble solids build-up on the column. The anionic radionuclides are not substantially removed in the column and pass with the coolant through the column. The anionic radionuclides can be subsequently removed by treatment with conventional anion exchange resins. Upon regeneration of the conventional anion exchange resin after it becomes loaded, the regenerant solution containing the anionic radionuclides can be concentrated by evaporation and the resulting concentrate can be molecularly stuffed pursuant to the procedures described in U.S. Pat. No. 4,110,096 into the pores of the porous glass in the ion exchange column after said porous glass had become substantially loaded with silicon-bonded radionuclide cation oxy groups. It is preferred to first dry the loaded porous glass so that the anionic radionuclide concentrate can readily enter the pores of the porous glass. The anionic radionuclides can be precipitated or deposited within the pores of the porous glass by the careful drying procedures disclosed in U.S. Pat. No. 4,110,096. Thereafter, columns containing the porous glass particles can be heated to drive off volatiles, to decompose decomposables and drive off non-radioactive decomposition products, to collapse the pores of the particles and sinter same into a unitary mass and to collapse the Vycor glass column around the sintered mass thereby enveloping the filtered solids and the sintered mass glass particles containing the cationic and anionic radinuclides within the collapsed Vycor glass column. While the glass column cracks because of differential thermal contraction it still contains and further immobilizes the radioactive materials and forms a product that is many times more durable than cement or metal drum presently in use. There is thus provided a durable package of concentrated radionuclides which is highly resistant to leaching by water or other fluids. As illustrated in Example 24, liquid radwaste that must be satisfactorily treated and disposed of can be highly dilute. The volume of dilute radwaste treated with a given amount of ion exchange porous glass pursuant to Example 24 can be practically unlimited before all the available exchange sites (i.e. silicon-bonded sodium oxy groups) in the porous silicate glass are filled by radioactive cations. For example, the weight of the dilute liquid radwaste described in Example 24 that could be expected to be treated befoe exhausting all exchange sites would be of the order of 10.sup.9 or more times the weight of the ion exchange porous glass employed. Furthermore, it could be expected that other parts of the system would require overhaul, e.g., repair or replacement of pumps or piping or other equipment, before the ion exchange silicate glass becomes exhausted. Consequently, it is quite possible, if not probable, that the radioactivity of the resulting porous glass when sintered for storage may never reach 1 millicurie or even 1 microcurie per cc. of the glass. In the absence of malfunction requiring overhaul of the other parts of the radwaste treatment system, 100 or less to 10.sup.9 or more, preferably 100 to 10.sup.6, weight parts of radwaste can be treated for each weight part of porous silicate glass within the Vycor tube ion exchange column. EXAMPLE 25 Use of Porous Powder in Non-Porous Tube to Encapsulate A non-radioactive nitrate mixture was used to similate the United Kingdom UKM-22 commercial waste whose composition is reported in terms of oxides in Table 5. Various amounts of nitrates were mixed together in such a proportion as to yield the appropriate oxide concentrations given in Table 5. Appropriate amounts of nitrates whose total weight corresponds to a total of 2 g oxides were placed in a 250 ml beaker; 20 ml H.sub.2 O was added; the solution was stirred and heated up slowly to 80.degree. C. at which temperature a light brown solution containing some undissolved salts was obtained. 18 g of porous glass prepared as in Example 3 was then added to the solution as to give a 10% loading of waste oxides with respect to the final glass. The volume ratio of solution to glass powder was close to 1:1. The mixture was dried at 90.degree. C. Approximately 3 g of the dry mixture was heated under vacuum in a Vycor tube similar to the one described in Example 5 according to the following schedule: ______________________________________ Time (hour:minute) T (.degree.C.) Pressure, m Torrs ______________________________________ 9:45 AM 25 25 10:15 AM 65 30 11:15 AM 278 26 11:30 AM 342 38 11:40 AM 383 32 11:50 PM 403 68 12:05 PM 520 44 3:20 PM 1300 36 3:45 PM 1310 16 4:15 PM 1310 16 ______________________________________ The finished glass product showed that the pores of the powder and the grains inside the tube were all sintered. In addition, the tube was completely collapsed but cracked during air quenching. The finished product was powdered to increase its surface area and was subjected to a leaching test at pH 5.6 and at 70.degree. C. for various exposure times. The results as reported in Table 6 show that the glass sample possesses an excellent chemical durability. TABLE 5 ______________________________________ United Kingdom, UKM-22 composition. Reported Simulated Reported Simulated Oxide wt % wt % Oxide wt % wt % ______________________________________ Al.sub.2 O.sub.3 19.89 19.89 ZrO.sub.2 5.57 5.57 Rb.sub.2 O 0.43 0.43 PO.sub.4 0.93 0.93 Cs.sub.2 O 3.00 3.00 Cr.sub.2 O.sub.3 2.18 2.18 MgO 24.68 24.68 MoO.sub.3 6.89 6.89 SrO 1.25 1.25 Fe.sub.2 O.sub.3 10.63 10.63 BaO 1.48 1.48 RuO.sub.2 2.65 2.65 Y.sub.2 O.sub.3 0.66 0.66 NiO.sub.2 1.40 1.40 La.sub.2 O.sub.3 1.71 1.71 PdO 1.71 1.71 Pr.sub.6 O.sub.11 1.67 -- ZnO 1.71 1.71 Nd.sub.2 O.sub.3 7.08 7.08 U.sub.3 O.sub.8 0.23 Replaced by CeO.sub.2 CeO.sub.2 3.85 3.85 SO.sub.4 0.39 0.39 ______________________________________ TABLE 6 ______________________________________ Chemical Durability Of Product Obtained In Example 25 In Deionized Water Having An Initial pH of 5.6 Glass Component and Leach Rate Sample SiO.sub.2 Ln* Fe Na Cs Sr ______________________________________ Core and Clad 295 32 <1 <4 <20 <1 Powdered Core Powdered 127 42 11 17 3 8 ______________________________________ Data taken between Day 12 and Day 15, 70.degree. C., 71 hrs. Leach rates are in ng of waste dissolved per cm.sup.2 of surface area o powdered product per day. *Includes all lanthanides. The leach rates reported in Tables 7 and 9 below have been normalized by the amount of the component present in the glass. Thus, they represent the leach rate the glass would have it the measurement was made only on that component. The glass is dissolving at the silica leach rate. The sodium, strontium and cesium diffuse to the surface and are initially leached at a faster rate. Iron and lanthanites concentrate at the surface. Eventually, the whole glass will leach at the silica rate. EXAMPLE 26 Use of Porous Powder in Non-Porous Tube To Encapsulate A non-radioactive nitrate mixture similar to the one described in Example 25 to simulate the UKM-22 waste was prepared. However, in the preparation of this nitrate mixture, Zr(NO.sub.3).sub.4 and K.sub.2 MoO.sub.4 was dissolved separately from the other nitrates with sufficient amount of concentrated HNO.sub.3, the others being dissolved in a 3MHNO.sub.3 solution or in water. The two solutions were then mixed together and no precipitate was observed. Phosphoric acid and sulfuric acid were subsequently added to the solution to yield appropriate amounts of PO.sub.4.sup..tbd. and SO.sub.4.sup..tbd.. A white gelatinous precipitate appeared and did not dissolve upon heating up to 70.degree. C. About 50% of the nitrates precipitated out when the solution was evaporated down to about 15 ml. Eight grams of porous glass prepared as in Example 3 were then added to the solution to give a 20% loading of waste oxides with respect to the final glass. The volume ratio of solution to glass powder was about 1:1. The mixture was dried at 90.degree. C. for about 16 hrs. Approximately 3 g of the dry mixture was placed under vacuum in a Vycor tube having an outside diameter of 13 mm and a wall thickness of 1.5 mm. The mixture was heated to 600.degree. C. at 50.degree. C./hr. After holding at 600.degree. C. for 48 hrs, the tube was subjected to a temperature jump to 1240.degree. C. where the pores and the grains inside the tube were well sintered. The tube, however, did not collapse and bubbles were formed in the waste-glass matrix. Moreover, the tube cracked during air quenching. Leaching tests were performed on the core of the sample. The results reported in Table 7 show that it has an excellent chemical durability. TABLE 7 ______________________________________ Chemical Durability Of Product Obtained In Example 26 In Deionized Water Having An Initial pH of 5.6 Time Glass Component and Leach Rate (Days) SiO.sub.2 Fe Ln* Na Sr Cs ______________________________________ 0.34 6,190 1150 737 3.61 .times. 10.sup.5 3,260 <1000 1.3 963 120 344 <2,500 6,340 300 2.2 550 30 400 <2,500 2,200 <300 3.3 370 49 550 <2,500 2,300 1,400 5.7 200 12 <80 <2,500 1,400 120 9.3 260 <13 50 <2,500 680 <320 12.2 220 3 210 <2,500 1,900 150 15.2 230 <13 56 -- 2,000 <320 ______________________________________ Data taken at 70.degree. C. Leach rates are in ng of waste dissolved per cm.sup.2 of surface area o powdered product per day. *Includes all lanthanites. EXAMPLE 27 Use of Porous Powder In Ion-Exchanged Tube To Encapsulate A mixture containing non-radioactive nitrates and porous glass was prepared as in Example 26 but with only 5% loading of oxides with respect to the final glass. Approximately 3 g of the dry mixture was introduced in an ion-exchanged tube which was prepared as follows: an opened porous tube having an outside diameter of 10 mm, a wall thickness of .about.1 mm and a length of 20 cm was prepared as in Example 2. The porous tube was then soaked in a solution containing 200 ppm Cs with enough NH.sub.4 OH to give a pH of 10 for 18 hrs, and washed in room temperature water until a pH of 7 was obtained. The Cs exchanged tube was subsequently dried under vacuum and was heated from room temperature to 600.degree. C. at 15.degree. C./hr and from 600.degree. C. to 870.degree. C. at 50.degree. C./hr to collapse the pores. One end of the tube was then sealed using a torch prior to the introduction of the mixture of simulated wastes and porous glass. The mixture was then heated under vacuum in the tube according to the following schedule: ______________________________________ Time (hour:minute) T (.degree.C.) Pressure, m Torrs ______________________________________ 11:00 AM 22 15 11:20 AM 180 100 11:25 AM 200 100 11:35 AM 252 50 12:02 PM 330 48 12:10 PM 470 36 12:53 PM 547 28 1:00 PM 775 25 1:20 PM 875 24 1:30 PM 927 24 1:35 PM 1010 24 1:47 PM 1075 24 2:00 PM 1100 24 ______________________________________ The finished glass article showed that the collapsing of the tube was complete and there were no cracks. The grains inside the tube, however, did not completely sinter. Here the thermal expansion coefficients of the tube and powder were matched. However, complete sintering was not achieved because the collapsing temperature of the tube (about 1100.degree. C.) was too low for the nuclear waste composition and loading level utilized. The composition of the ion exchange tube was measured to be 0.5 weight percent Cs. EXAMPLE 28 Use of Porous Powder In Non-Porous Tube To Encapsulate A non-radioactive nitrate mixture was used to simulate the West-Valley PW-8a waste whose composition is reported in terms of oxides in Table 8. Various amounts of nitrates were first dissolved separately in 3 M HNO.sub.3 or in water and then were mixed in such a proportion as to yield the appropriate oxide concentrations given in Table 8. The solution containing appropriate amounts of nitrates plus some undissolved salts whose total weight corresponds to a total of 4 g oxides was evaporated down to near dryness and was then mixed with 16 g of porous glass prepared as in Example 3 as to yield a loading of 20% waste oxides with respect to the final glass. The mixture was subsequently dried at 90.degree. C. Approximately 3 g of the dry mixture was heated under vacuum in a Vycor tube similar to the one described in Example 5. The mixture was heated to 600.degree. C. then was subjected to a temperature jump to about 1250.degree. C. at which temperature the waste porous glass mixture sintered completely. The Vycor tube did not fully collapse and cracked during air quenching. Leaching tests were performed on the core of the sample. The results reported in Table 9 show that it has an excellent chemical durability. TABLE 8 ______________________________________ West-Valley PW-8a Composition Reported Simulated Reported Simulated Oxide wt % wt % Oxide wt % wt % ______________________________________ Na.sub.2 O 16.62 16.62 TeO.sub.2 0.86 -- Fe.sub.2 O.sub.3 34.29 34.29 Cs.sub.2 O 1.14 1.14 Cr.sub.2 O.sub.3 1.36 1.36 BaO 1.85 1.85 NiO 1.74 1.74 Y.sub.2 O.sub.3 0.05 0.05 P.sub.2 O.sub.5 1.58 1.58 La.sub.2 O.sub.3 6.05 6.05 Rb.sub.2 O 0.21 0.21 CeO.sub.2 12.09 12.09 SrO 1.25 1.25 Pr.sub.6 O.sub.11 1.06 1.06 ZrO.sub.2 5.84 5.84 Nd.sub.2 O.sub.3 3.62 3.62 MoO.sub.3 7.54 7.54 Sm.sub.2 O.sub.3 0.64 0.64 Rh.sub.2 O.sub.3 0.36 0.36 Eu.sub.2 O.sub.3 0.17 0.17 Ag.sub.2 O 0.104 0.104 Gd.sub.2 O.sub.3 0.43 0.43 CdO 0.15 0.15 ______________________________________ 9 ______________________________________ Chemical Durability Of Product Obtained In Example 28 In Deionized Water Having An Initial pH of 5.6 Time Glass Component and Leach Rate (Days) SiO.sub.2 Fe Ln* Na Sr Cs ______________________________________ 0.34 2800 62 <32 6500 560 223 1.3 905 8 370 2500 2000 <630 2.2 550 25 440 1400 240 370 3.25 430 12 440 1360 1200 670 5.7 200 100 150 870 880 340 9.3 280 <25 150 780 600 630 12.2 313 <1 200 780 780 770 15.2 300 3 120 840 -- 620 ______________________________________ Data taken at 70.degree. C. Leach rates are in ng of waste dissolved per cm.sup.2 of surface area o powdered product per day. **Includes all lanthanites. EXAMPLE 29 Use Of Porous Powder In Ion-Exchanged Tube To Encapsulate The porous powder mixed with nuclear waste described in Example 28 was used in a tube made according to Example 27. The mixture was heated in vacuum to 600.degree. C. then subjected to a temperature jump to about 1100.degree. C. at which temperature the waste porous glass mixture sintered conpletely. The ion exchanged tube did collapse completely. However, it cracked during air quenching. Upon examination of the core material it was found that it had been completely sintered and that it was a good quality glass. Thus, by increasing the loading level of nuclear waste from Example 27, we were able to lower the sintering temperature to below the collapsing temperature of the ion exchanged tube. However, we put an excessive amount of nuclear waste in this experiment and the expansion coefficient was slightly too high causing a small number of cracks. To achieve a completely sintered uncracked product with ion exchanged tubes used in Example 27 and 29, intermediate loading levels should be used. For example, loading levels between 8 and 12%.
	claims	1. A method for evaluating optical performance of an inspection apparatus that uses a charged particle beam, the method comprising:providing the inspection apparatus with a charged-particle-beam (CPB) mapping optical system and an electron emitter that emits an electron beam spontaneously;providing, on a surface of the electron emitter, an evaluation chart having a pattern over the surface;using the CPB mapping optical system, producing an image of the evaluation chart; andbased on an observation of the image of the evaluation chart, determining a resolution of the CPB mapping optical system. 2. A method for evaluating optical performance of an inspection apparatus that uses a charged particle beam, the method comprising:providing the inspection apparatus with a charged-particle-beam (CPB) mapping optical system and an electron emitter that emits an electron beam spontaneously;providing, on a surface of the electron emitter, an evaluation chart having a dot pattern over the surface;using the CPB mapping optical system, producing an image of the evaluation chart; andbased on an observation of the image of the evaluation chart, adjusting an optical axis of the mapping optical system. 3. A method for evaluating optical performance of an inspection apparatus that uses a charged particle beam, the method comprising:providing the inspection apparatus with a charged-particle-beam (CPB) optical system and an electron emitter that emits an electron beam spontaneously;providing, on a surface of the electron emitter, an evaluation chart having a dot pattern over the surface;using the CPB optical system, observing the evaluation chart; andbased on the observation of the evaluation chart, adjusting an optical axis of the CPB optical system. 4. A method for evaluating optical performance of an inspection apparatus that uses a charged particle beam, the method comprising:providing the inspection apparatus with a charged-particle-beam (CPB) mapping optical system and an electron emitter that emits an electron beam spontaneously;providing, on a surface of the electron emitter, an evaluation chart having a cross pattern or an L-shaped pattern over the surface;using the CPB mapping optical system, producing an image of the evaluation chart; andbased on an observation of the image of the evaluation chart, adjusting an aberration of the CPB mapping optical system.
042971683	description	DETAILED DESCRIPTION OF THE INVENTION In the preferred practice of this invention, the additive is associated with the fuel in any suitable manner as by mechanically blending the additive in power form with the nuclear fuel material in a similar finely-divided condition. It is also feasible, according to this invention, to apply the additive as a coating to part or all of the surface of a fuel pellet, or it may be applied as a coating on the inside surface of the cladding for contact with fuel pellets loaded therein. As indicated above, it is also contemplated that the additive in powder form can be disposed in the pellet assembly as it is loaded into cladding. In any event, it is desirable that the vanadium oxide additive be distributed in respect to the nuclear fuel material to insure that substantially all cadmium generated as a fission product in the operation of the reactor comes into contact with and is reacted with the additive during reactor operation so as to obtain the new results and advantages of this invention described above. Generally, the new and highly useful cadmium immobilization result of this invention can be achieved in accordance with relatively small amounts of V.sub.2 O.sub.4 or V.sub.2 O.sub.5. Thus, 0.16 gram of V.sub.2 O.sub.5 is sufficient for the present purpose in a reactor operation at 20,000 megawatt days per metric ton of uranium in a boiling water reactor fuel rod which generated 0.11 gram of cadmium. In the best practice presently contemplated, the vanadium oxide additive will be present in association with the fuel in one or the other of the several alternative ways described above in such stoichiometric amount. Appreciably less than such stoichiometric amount will leave the way open to some extent for cadmium embrittlement of cladding, while use of substantially more than the stoichiometric amount burdens the system with inert material, using space that should be occupied by fissile or fertile material. When the additive is incorporated in the fuel elements, they make take any desired geometric form or configuration, but it is preferred that the nuclear fuel material be in the form of right cylindrical pellets which are incorporated in a tubular cladding of a zirconium alloy. The swelling of the pellets in the cladding is accommodated by providing porosity in the fuel pellet or by forming it with dished ends or axial openings or the like to accommodate such swelling. From the foregoing description, it will be understood that this invention achieves the chemical inerting of reactive fission product cadmium through the use of V.sub.2 O.sub.4 or V.sub.2 O.sub.5 additives which react with cadmium under normal nuclear reactor operating conditions to form stable compounds so that fission product cadmium is not available or free to react with or attack fuel cladding or any other metal that it may come in contact with during reactor operation. In this manner, the additive which is effective for the purposes of this invention blocks potential cladding-fission product reaction and so increases the cladding reliability and useful life. In a test conducted for the purpose of confirming that cadmium embrittles zirconium alloy cladding material under elevated temperature conditions, a Zircaloy-2 tensile test specimen was broken in argon at 300.degree. C. after undergoing a 75 percent reduction in cross-sectional area and with a plastic strain of about 15 percent following a maximum stress of 60,000 psi. Fracture morphology was ductile. Then in a repetition of that test but for the presence of cadmium in contact with the test specimen, breakage occurred as a transgranular cleavage fracture with zero reduction in area and zero plastic strain at maximum stress of 40,000 psi before reaching the yield point of the specimen. Many incipient cracks were observed in the specimen on conclusion of the test. Similar results to those of the latter test were obtained in subsequent tests performed in the same manner but at temperatures between 250.degree. C. and 350.degree. C. involving the use of solid cadmium (below 321.degree. C.), liquid cadmium(above 321.degree. C.) and cadmium dissolved in liquid cesium at temperatures both above and below 321.degree. C. In testing the basic new cadmium inerting concept of this invention, oxides of variable-valence metals were equilibrated with cadmium at 350.degree. C. in evacuated quartz capsules in a thermal gradient furnace. Either a reaction occurred or it did not; and where the test result was positive in this sense, the compounds formed were stable up to the approximately 1000.degree. C. temperature limit of the furnace. As indicated above, V.sub.2 O.sub.4 and V.sub.2 O.sub.5 did so react under these conditions in this test with the apparent formation of cadmium oxide and a lower oxide of vanadium. No such reaction was observed in tests of this kind involving the use of either TiO.sub.2 or Nb.sub.2 O.sub.5, CeO.sub.2 or depleted UO.sub.2. In out-of-pile experiments performed with V.sub.2 O.sub.5, it was found that 0.1 gram of cadmium was immobilized or gettered by 1.6 grams of V.sub.2 O.sub.5 at temperatures between 300 and 950.degree. C. (the maximum test temperature). Actually, on visual examination it was observed that only about one-tenth of the total volume of V.sub.2 O.sub.5 was darkened (i.e., changed in color from yellow to black), indicating that a reaction of one-to-one stoichiometry had taken place. In using V.sub.2 O.sub.4 or V.sub.2 O.sub.5 or mixtures of them in accordance with this invention to fill the gap between the nuclear fuel and the cladding of a fuel rod, the vanadium oxide material in powder form may be packed lightly in place. With the volume of that gap typically being about 14.5 cc, a gap-filling load would be about 11.6 grams, which would insure inerting of the cadmium released at all locations in the fuel rod during reactor operation. When it is desired to provide the embrittlement protection of this invention in locations between fuel pellets, a 5-mil layer of V.sub.2 O.sub.5, for example, may be disposed between each pair of pellets. Thus, in a typical fuel rod assembly of 100 fuel pellets, each of 0.87 square centimeters end surface area, a total of about 0.9 gram of V.sub.2 O.sub.5 will be incorporated in the fuel rod. As indicated above, this amount will be in substantial excess of the stoichiometric equivalent of the cadmium produced in the normal useful life of the fuel rod in the typical boiling water nuclear reactor operation.
	abstract	The present invention provides method and apparatus for automatically correcting aberrations in a charged-particle beam. The apparatus has a memory for storing image data obtained by scanning a specimen with the beam. A four-sided region-blurring device reads the image data from the memory and blurs regions close to the four sides of an image represented by the image data. A probe profile extractor extracts the probe profile from the image that has been blurred as mentioned above. A correction amount-calculating unit performs extraction of amounts of features, calculations of aberrations. A correcting unit corrects the aberration corrector.
	description	This invention was made with government support under Contract Number DE-AC07-05-ID14517 awarded by the United States Department of Energy. The government has certain rights in the invention. Embodiments of the disclosure relate generally to electrochemical cells for reducing one or more oxides, and related methods. More particularly, embodiments of the disclosure relate to the direct oxide reduction (DOR) electrochemical cells for reducing one or more metal oxides at the cathode of the electrochemical cell, the electrochemical cell comprising one or more anode materials that are substantially inert in a molten salt electrolyte of the electrochemical cell, and related methods of reducing metal oxides in the electrochemical cell. Processing of spent nuclear fuels is often performed with the goal of extracting uranium and plutonium from the spent nuclear fuels. One example of processing spent nuclear fuels is the so-called “Purex” process wherein the spent nuclear fuel is dissolved in nitric acid. Insoluble solids of the spent nuclear fuel are removed from the nitric acid solution. An organic solvent (e.g., tributyl phosphate (TBP)) mixed with a hydrocarbon solvent, such as kerosene, is used to extract the uranium and plutonium from the nitric acid solution to form UO2(NO3)2·2TBP complexes and similar complexes of plutonium. The plutonium is subsequently separated from the uranium, such as by exposure to aqueous ferrous sulphamate. Other methods of processing used oxide nuclear fuels include dissolving the used oxide nuclear fuel in an electrolyte and electrowinning the dissolved materials (e.g., dissolved uranium ions) by passing a current through the electrolyte including the dissolved materials. The current forces the dissolved uranium ions to move toward the cathode, where the uranium ions are reduced and deposited on the cathode surface. In addition to nuclear fuels, many metal oxides may be purified and/or converted to their constituent metals in an electrochemical cell. Reduction of metal oxides in an electrochemical cell conventionally involves dissolution of the metal oxide in an electrolyte. The electrolyte of such electrochemical cells may include a molten salt electrolyte. Conventional anode materials used in such electrochemical cells include graphite, tin oxide, cermets, ceramic materials, and metals such as platinum. The molten salt is often corrosive to one or more portions of the electrochemical cell, such as the anode. For example, molten salts that contain lithium may degrade platinum-containing anodes by way of forming lithium platinate (Li2PtO3). A portion of the lithium platinate dissolves in the molten salt electrolyte. Over time, the anode material is thinned and eventually dissolves completely in the molten salt electrolyte or otherwise becomes unusable in the electrochemical cell. In addition, depending on the particular configuration of the electrochemical cell, oxygen gas may evolve at the anode. The evolved oxygen may react with the platinum anode to form oxides of platinum (e.g., platinum (II) oxide (PtO), platinum (IV) oxide (PtO2), Pt3O4) that dissolve in the molten salt electrolyte. Since the oxides of platinum are soluble in the molten salt electrolyte, the anode material is dissolved in the molten salt electrolyte and, as a result, the anode material is thinned, further reducing the usable life of the anode. In addition, spent nuclear fuels include fission byproducts, which often include corrosive gases such as selenium, tellurium, and iodine. Under operating conditions of the electrochemical cell, these corrosive gases undesirably react with the anode to form soluble intermetallic compounds. As the soluble intermetallic compounds form and dissolve in the molten salt, the anode is further degraded. Over time, as more platinum reacts with the components of the molten salt (e.g., lithium oxide ions, and corrosive gases), the platinum anode material is progressively thinned and consumed in the electrochemical cell. As another example, graphite-containing anodes may react with the oxide ions in the electrochemical cell, generating carbon dioxide and carbon monoxide, as well as carbon dusts on the surface of the molten salt electrolyte. The carbon dust contaminates the components of the electrochemical cell. The generation of carbon dioxide, carbon monoxide, and carbon dust consumes the graphite anode. As the anode thins and is consumed in the electrochemical cell, long term operation of the electrochemical cell is hindered. If sufficient portions of the anode material are consumed, the anode needs to be replaced for the desired electrolytic reduction reaction to proceed. Embodiments disclosed herein include methods of reducing metal oxides with a molten salt electrolyte and to related methods. For example, in accordance with one embodiment, a method of reducing spent nuclear fuels comprising uranium oxide comprises providing an electrochemical cell comprising a working electrode, and a counter electrode comprising one or more materials selected from the group consisting of osmium, ruthenium, rhodium, iridium, palladium, silver, gold, lithium iridate, lithium ruthenate, lithium rhodates, a lithium tin oxygen compound, a lithium manganese compound, strontium ruthenium ternary compounds, calcium iridate, strontium iridate, calcium platinate, strontium platinate, magnesium ruthenate, magnesium iridate, sodium ruthenate, sodium iridate, potassium iridate, and potassium ruthenate, disposing a uranium oxide material on or proximate the working electrode, exposing the uranium oxide to a molten salt electrolyte comprising at least one of lithium chloride, lithium oxide, calcium chloride, calcium oxide, and sodium chloride, and providing an electric current between the counter electrode and the working electrode to reduce the uranium oxide material. In additional embodiments, a method of direct oxide reduction comprises forming a molten salt electrolyte in an electrochemical cell, disposing at least one metal oxide in the electrochemical cell, disposing a counter electrode comprising a material selected from the group consisting of osmium, ruthenium, rhodium, iridium, palladium, platinum, silver, gold, lithium iridate, lithium ruthenate, a lithium rhodate, a lithium tin oxygen compound, a lithium manganese compound, strontium ruthenium ternary compounds, calcium iridate, strontium iridate, calcium platinate, strontium platinate, magnesium ruthenate, magnesium iridate, sodium ruthenate, sodium iridate, potassium iridate, and potassium ruthenate in the electrochemical cell, and applying a current between the counter electrode and the at least one metal oxide to reduce the at least one metal oxide. In further embodiments, an electrochemical cell comprises a counter electrode comprising a material selected from the group consisting of osmium, ruthenium, rhodium, iridium, palladium, platinum, silver, gold, lithium iridate, lithium ruthenate, a lithium rhodate, a lithium tin oxygen compound, a lithium manganese compound, strontium ruthenium ternary compounds, calcium iridate, strontium iridate, calcium platinate, strontium platinate, magnesium ruthenate, magnesium iridate, sodium ruthenate, sodium iridate, potassium iridate, and potassium ruthenate, a working electrode comprising a metal oxide to be reduced in the electrochemical call, and a molten salt electrolyte comprising a molten salt comprising at least one of an alkali halide salt and an alkaline earth metal halide salt. Illustrations presented herein are not meant to be actual views of any particular material, component, or system, but are merely idealized representations that are employed to describe embodiments of the disclosure. The following description provides specific details, such as material types, dimensions, and processing conditions in order to provide a thorough description of embodiments of the disclosure. However, a person of ordinary skill in the art will understand that the embodiments of the disclosure may be practiced without employing these specific details. Indeed, the embodiments of the disclosure may be practiced in conjunction with conventional fabrication techniques employed in the industry. In addition, the description provided below does not form a complete process flow, apparatus, system or method for reducing a metal oxide. Only those process acts and structures necessary to understand the embodiments of the disclosure are described in detail below. Additional acts to reduce a metal oxide may be performed by conventional techniques. Also note, any drawings accompanying the present application are for illustrative purposes only, and are thus not drawn to scale. Additionally, elements common between figures may retain the same numerical designation. As used herein, the term “platinum group metal” (PGM) means and includes a metal including at least one of ruthenium, osmium, rhodium, iridium, palladium, and platinum. According to embodiments described herein, an electrochemical cell comprises a molten salt electrolyte, a counter electrode (e.g., an anode) in contact with the molten salt electrolyte, and a working electrode (e.g., a cathode) in contact with the molten salt electrolyte. The molten salt electrolyte includes a molten salt of an alkali metal halide salt, an alkaline earth metal halide salt, an alkali metal oxide, an alkaline earth metal oxide, or combinations thereof. The working electrode may comprise a metal oxide to be reduced in the electrochemical cell or the metal oxide to be reduced may be in direct contact with the working electrode. In some embodiments, the working electrode consists essentially of the metal oxide to be reduced. The metal oxide may comprise a transition metal oxide, a lanthanide oxide, an actinide oxide, or combinations thereof. In some embodiments, the metal oxide comprises spent uranium oxide nuclear fuel (e.g., depleted uranium oxide fuel), unirradiated nuclear fuel (e.g., enriched uranium oxide fuel), or combinations thereof. In use and operation, the metal oxide is reduced at the working electrode and oxide ions are generated at the working electrode. The reduced metal remains at the working electrode and the oxide ions are dissolved in the molten salt electrolyte. Accordingly, the metal is not substantially dissolved in the molten salt electrolyte. Responsive to exposure to an electric current between the counter electrode and the working electrode, the oxide ions move from the working electrode to the counter electrode through the molten salt electrolyte. The counter electrode comprises a material that is substantially inert or otherwise not substantially consumed in the electrochemical cell. In some embodiments, the counter electrode comprises a material that is stable in an oxidizing atmosphere at an operating temperature of the electrochemical cell. In some embodiments, the material of the counter electrode comprises at least one platinum group metal (PGM), at least one precious metal (e.g., gold or silver), or a combination thereof. By way of nonlimiting example, the counter electrode is selected from the group consisting of at least one of osmium, ruthenium, rhodium, iridium, palladium, platinum, silver, gold, lithium iridate (Li2IrO3), lithium ruthenate (Li2RuO3), a lithium rhodate (LiRhO2, LiRhO3), a lithium tin oxygen compound (e.g., Li2SnO3), a lithium manganese oxygen compound (e.g., Li2MnO3), calcium ruthenate (CaRuO3), strontium ruthenium ternary compounds (e.g., SrRuO3, Sr2RuO3, Sr2RuO4), CaIrO3, strontium iridate (e.g., SrIrO3, SrIrO4, Sr2IrO4), calcium platinate (CaPtO3), strontium platinate (SrPtO4), magnesium ruthenate (MgRuO4), magnesium iridate (MgIrO4), sodium ruthenate (Na2RuO4), sodium iridate (Na2IrO3), potassium iridate (K2IrO3), and potassium ruthenate (K2RuO4). In some embodiments, the counter electrode comprises a substrate comprising a different material than the at least one platinum group metal or the at least one precious metal and the substrate is coated with the material of the counter electrode. The counter electrode may be substantially inert in the electrochemical cell. The counter electrode may resist attack from the molten salt electrolytes, which may be corrosive at high temperatures (e.g., greater than about 600° C., greater than about 800° C., etc.) under oxidizing conditions. The counter electrode may exhibit good electrical conductivity suitable for operation in the electrochemical cell. Accordingly, the material of the counter electrode may not be consumed during the electrochemical reduction reaction (e.g., direct oxide reduction) and the counter electrode may not need to be replaced as in conventional electrochemical cells. FIG. 1 is a simplified schematic of a system 100 including an electrochemical cell 102 for reducing one or more metal oxides to form one or more metals. The electrochemical cell 102 may be configured as a so-called “direct oxide reduction” (DOR) electrochemical cell. In other words, the electrochemical cell 102 may be configured to reduce one or more oxides. The electrochemical cell 102 may be contained within a gas-tight enclosure 104, which may include an inlet 106 and an outlet 108. The inlet 106 is configured for providing, for example, a gas to the enclosure 104 for maintaining a gas pressure within the enclosure 104. Gases may be removed from the enclosure 104 via the outlet 108. In some embodiments, the gas comprises an inert gas, such as argon, helium, or a combination thereof. The enclosure 104 may include a furnace or other heating element for heating or maintaining a temperature of a molten salt electrolyte 110 in the electrochemical cell 102. Although FIG. 1 illustrates that the enclosure 104 include the inlet 106 and the outlet 108, the disclosure is not so limited. In other embodiments, the enclosure 104 may be configured as a so-called “glove box” wherein the enclosure is not configured with an inlet 106 and an outlet 108 for gas flow into and out of the electrochemical cell 102 during operation thereof. The electrochemical cell 102 may include a crucible 112 comprising a metal, glassy carbon, ceramic, a metal alloy, or another material. In some embodiments, the crucible 112 comprises a non-metallic material, such as alumina (Al2O3), magnesia (MgO), glass carbon, graphite, boron nitride, another material, or combinations thereof. In other embodiments, the crucible comprises a metal or metal alloy, such as, for example, nickel, molybdenum, tantalum, stainless steel, alloys of nickel and copper, alloys of nickel, chromium, iron, and molybdenum, alloys of nickel, iron, and molybdenum, and combinations thereof. The molten salt electrolyte 110 may be disposed in the crucible 112. The electrochemical cell 102 may further include at least one counter electrode 114 (which may also be referred to as an anode) and at least one working electrode 116 (which may also be referred to as a cathode). In some embodiments, the electrochemical cell 102 further includes a reference electrode 118 configured for monitoring a potential in the electrochemical cell 102. In some embodiments, a sheath 122 is disposed around at least a portion of one or more of the counter electrode 114, the working electrode 116, and the reference electrode 118. The sheath 122 may be configured to provide electrical insulation between the respective electrodes and the crucible 112. In some embodiments, the sheath 122 comprises alumina (e.g., an alumina tube), magnesia, or a combination thereof. The reference electrode 118 may be in electrical communication with the counter electrode 114 and the working electrode 116 and may be configured to monitor the potential difference between the counter electrode 114 and the working electrode 116. Accordingly, the reference electrode 118 may be configured to monitor the cell potential of the electrochemical cell 102. The reference electrode 118 may include nickel, nickel/nickel oxide, glassy carbon, silver/silver chloride, one or more platinum group metals, one or more precious metals (e.g., gold), or combinations thereof. In some embodiments, the reference electrode 118 comprises glassy carbon. In other embodiments, the reference electrode 118 comprises nickel, nickel oxide, or a combination thereof. In yet other embodiments, the reference electrode 118 comprises silver/silver chloride. A potentiostat 124 may be electrically coupled to each of the counter electrode 114, the working electrode 116, and the reference electrode 118. The potentiostat 124 may be configured to measure and/or provide an electric potential between the counter electrode 114 and the working electrode 116. The difference between the electric potential of the counter electrode 114 and the electric potential of the working electrode 116 may be referred to as a cell potential of the electrochemical cell 102. The system 100 may be configured to reduce one or more metal oxides to a substantially pure metal (e.g., a metal in a substantially unoxidized state) or a metal alloy. In some such embodiments, the working electrode 116 includes at least one oxide (e.g., at least one metal oxide) to be reduced in the electrochemical cell 102. The working electrode 116 may be in electrical communication with a basket 120 configured to carry one or more metals to be reduced in the electrochemical cell 102. The basket 120 may comprise nickel, cobalt, iron, molybdenum, stainless steel, alloys of nickel and copper, alloys of nickel, chromium, iron, and molybdenum, alloys of nickel, iron, and molybdenum, another material, or combinations thereof. In some embodiments, the basket 120 comprises nickel. In other embodiments, the electrochemical cell 102 does not include the basket 120 and the working electrode 116 comprises the metal oxide or a combination of metal oxides to be electrolytically reduced in the electrochemical cell 102. Stated another way, in some embodiments, the working electrode 116 comprises one or more metal oxides that are reduced to a metal (e.g., a substantially pure metal or a metal alloy) in the electrochemical cell 102. In some embodiments, the working electrode 116 consists essentially of the metal oxide, which may comprise one or more metals to be reduced. At least one of the working electrode 116 and the metal in the basket 120 may comprise a metal oxide. The metal oxide may comprise a transition metal oxide (such as a refractory metal oxide (e.g., titanium oxide, (TiO), titanium dioxide (TiO2), zirconium oxide (ZrO2), hafnium oxide (HfO2), vanadium oxide (V2O5), niobium oxide (NbO2, Nb2O5), tantalum pentoxide (Ta2O5), chromium oxide (CrO, Cr2O3, etc.), manganese oxide (MnO), nickel oxide (NiO), molybdenum oxide (MoO3), tungsten oxide (WO3, WO2), ruthenium oxide (RuO2), osmium oxide (OsO2, OsO4), rhodium oxide (Rh2O3), iridium oxide (IrO2)), iron oxide (Fe2O3, Fe3O4, etc.), cobalt oxide (CoO, Co2O3, Co3O4), nickel oxide (NiO, Ni2O3)), non-metal oxides (e.g., silicon dioxide (SiO2)), a lanthanide oxide (e.g., lanthanum oxide (La2O3)), cerium oxide (CeO2), neodymium oxide (Nd2O3), samarium oxide (Sm2O3), dysprosium oxide (Dy2O3), another oxide of a lanthanide element), an actinide oxide (e.g., actinium oxide (Ac2O3), thorium oxide (ThO2), uranium oxide (e.g., UO2), an oxide of another actinide element), or combinations thereof. In some embodiments, the metal oxide comprises an unirradiated nuclear fuel, such as enriched uranium oxide. In other embodiments, the metal oxide comprises a spent nuclear fuel, such as spent uranium oxide (e.g., UO2, U3O8, or a combination thereof). In some embodiments, the metal oxide comprises an oxide of more than one metal. Reduction of such oxides may form a metal alloy comprising the constituent metals of the metal oxides. In some embodiments, the metal oxide is disposed in the basket 120 and in electrical communication with the working electrode 116. In other embodiments, the working electrode 116 consists essentially of the metal oxide. The molten salt electrolyte 110 may include a material formulated and configured to facilitate reduction of the metal oxides. In some embodiments, the molten salt electrolyte 110 comprises an alkali halide salt, an alkaline earth metal halide salt, an alkali oxide, an alkaline earth metal oxide, or combinations thereof. By way of nonlimiting example, the molten salt electrolyte 110 may include lithium chloride (LiCl), lithium oxide (Li2O), sodium chloride (NaCl), calcium chloride (CaCl2), calcium oxide (CaO), lithium bromide (LiBr), potassium bromide (KBr), cesium bromide (CsBr), calcium bromide (CaBr2), potassium chloride (KCl), potassium bromide (KBr), strontium chloride (SrCl2), strontium bromide (SrBr2), or combinations thereof. In some embodiments, the molten salt electrolyte 110 comprises a eutectic mixture of sodium chloride and potassium chloride, and may further include calcium oxide. In some embodiments, the molten salt electrolyte 110 comprises lithium chloride and lithium oxide (LiCl—Li2O). In some such embodiments, the lithium oxide constitutes between about 1.0 weight percent (wt. %) and about 5.0 weight percent of the molten salt electrolyte 110, such as between about 1.0 weight percent and about 2.0 weight percent, between about 2.0 weight percent and about 3.0 weight percent, or between about 3.0 weight percent and about 5.0 weight percent of the molten salt electrolyte 110. The lithium chloride may constitute a remainder of the molten salt electrolyte 110. In some embodiments, the lithium oxide constitutes about 1.0 weight percent of the molten salt electrolyte 110. In other embodiments, the lithium oxide constitutes about 5.0 weight percent of the molten salt electrolyte 110. In some embodiments, the molten salt electrolyte 110 comprises lithium chloride and lithium oxide and the metal oxide comprises uranium oxide. In other embodiments, the metal oxide comprises one or more of manganese oxide, nickel oxide, and titanium monoxide. In other embodiments, the molten salt electrolyte 110 comprises calcium chloride and calcium oxide (CaCl2—CaO). In some such embodiments, the calcium oxide constitutes between about 1.0 weight percent and about 5.0 weight percent of the molten salt electrolyte 110, such as between about 1.0 weight percent and about 2.0 weight percent, between about 2.0 weight percent and about 3.0 weight percent, or between about 3.0 weight percent and about 5.0 weight percent of the molten salt electrolyte 110. The calcium chloride may constitute a remainder of the molten salt electrolyte 110. In some embodiments, the calcium oxide constitutes about 1.0 weight percent of the molten salt electrolyte 110. In other embodiments, the calcium oxide constitutes about 5.0 weight percent of the molten salt electrolyte 110. In some embodiments, the molten salt electrolyte 110 comprises calcium chloride and calcium oxide and the metal oxide comprises tantalum pentoxide, titanium oxide, a lanthanide oxide, an actinide oxide, or combinations thereof. The molten salt electrolyte 110 may be maintained at a temperature such that the molten salt electrolyte 110 is and remains in a molten state. In other words, the temperature of the molten salt electrolyte 110 may be maintained at or above a melting temperature of the molten salt electrolyte 110. By way of nonlimiting example, where the molten salt electrolyte 110 comprises lithium chloride and lithium oxide, the temperature of the molten salt electrolyte 110 may be between about 650° C. and about 700° C. Where the molten salt electrolyte 110 comprises calcium chloride and calcium oxide, the temperature of the molten salt electrolyte 110 may be between about 800° C. and about 950° C. Where the molten sale electrolyte 110 comprises sodium chloride and calcium chloride, the temperature thereof may be maintained between about 550° C. and about 950° C. However, the disclosure is not so limited and the temperature of the molten salt electrolyte 110 may be different than those described above. The molten salt electrolyte 110 may facilitate reduction of the metal oxide. In some embodiments, the metal oxide may be reduced at the cathode 116, according to Equation (1) below:MyOx(s)+ze−→y M+z/xO2− (1),wherein M is a metal (e.g., a transition metal, a lanthanide, an actinide, etc.), MyOx is the metal oxide, x is the stoichiometric amount of oxygen for the particular metal oxide, y is the stoichiometric amount of the metal in the metal oxide, and z is the stoichiometric amount of electrons for balancing the chemical reaction. The electrons are provided in the electrochemical cell 102 by provision of current to the working electrode 116, such as through the potentiostat 124. The oxide ions generated at the working electrode 116 may be transported from the working electrode 116 to the counter electrode 114 responsive to exposure to the applied electrical field (i.e., a polarization between the counter electrode 114 and the working electrode 116, provided by the potentiostat 124). The oxide ions may be oxidized at the counter electrode 114 according to Equation (2) below:2O2−→O2(g)+4e− (2).The oxygen gas generated at the counter electrode 114 may be evolved at the counter electrode 114. The electrons may be returned to the working electrode 116 surface. In use and operation, the metal oxide may be disposed in the electrochemical cell 102 and in contact with the molten salt electrolyte 110. An electric potential may be applied between the counter electrode 114 and the working electrode 116, providing a polarization field and a driving force for moving oxide ions dissolved from the metal oxide at the working electrode 116 to the counter electrode 114, facilitating reduction of the metal oxide at the working electrode 116. As described above, upon depositing the electrons at the counter electrode 114, the oxide anions may evolve as oxygen gas at the counter electrode 114. The counter electrode 114 may be formulated and configured to be substantially inert to the molten salt electrolyte 110, the oxide ions, and the oxygen gas. In addition, the counter electrode 114 may be substantially inert to gases that may evolve from the metal oxide. For example, where the metal oxide comprises a spent nuclear fuel, the metal oxide may include fission byproducts, such as selenium, tellurium, or iodine. The counter electrode 114 may be substantially inert to such gases. Accordingly, the counter electrode 114 may include a material formulated and configured to be substantially inert in the electrochemical cell 102 at operating conditions thereof. The counter electrode 114 may include a platinum group metal (PGM), a precious metal (e.g., silver or gold), or a combination thereof. By way of nonlimiting example, the counter electrode 114 may include ruthenium, osmium, rhodium, iridium, palladium, platinum, silver, gold, or combinations thereof. In some embodiments, the counter electrode 114 includes a ternary compound including a platinum group metal, oxygen, and one of an alkali metal and an alkaline earth metal. The ternary compound may have a general formula of M1M2Ox, wherein M1 is one or more of lithium, sodium, potassium, magnesium, calcium, or strontium, M2 is a platinum group metal such as ruthenium, osmium, rhodium, iridium, palladium, platinum, and combinations thereof, and x is an integer, such as 2, 3 or 4, depending on M1. In other embodiments, M2 may be tin or manganese. By way of nonlimiting example, the counter electrode 114 may comprise lithium iridate (Li2IrO3), lithium ruthenate (Li2RuO3), lithium rhodates (LiRhO2, LiRhO3), a lithium tin oxygen compound (e.g., Li2SnO3), a lithium manganese oxygen compound (Li2MnO3), calcium ruthenate (CaRuO3), strontium ruthenium ternary compounds (SrRuO3, Sr2RuO3, Sr2RuO4), calcium iridate (CaIrO3), strontium iridate (SrIrO3, SrIrO4, Sr2IrO4), calcium platinate (CaPtO3), strontium platinate (SrPtO4), magnesium ruthenate (MgRuO4), magnesium iridate (MgIrO4), sodium ruthenate (Na2RuO4), sodium iridate (Na2IrO3), potassium iridate (K2IrO3), and potassium ruthenate (K2RuO4), and combinations thereof. In some embodiments, where the counter electrode 114 comprises a ternary compound, M1 may be selected to comprise a material of the molten salt electrolyte 110. By way of nonlimiting example, where the molten salt electrolyte comprises lithium (e.g., LiCl/Li2O), M1 may be selected to be lithium. Similarly, where the molten salt electrolyte 110 comprises calcium (e.g., CaCl2/CaO), M1 may be selected to be calcium. In some embodiments, the counter electrode 114 comprises a monolithic material. In some such embodiments, the counter electrode 114 comprises a monolithic metal, such as monolithic iridium, monolithic ruthenium, monolithic osmium, monolithic rhodium, monolithic palladium, or monolithic platinum. In other embodiments, the counter electrode 114 comprises a monolithic structure of one of lithium iridate, lithium ruthenate, a lithium rhodate, a lithium tin oxygen compound, a lithium manganese compound, calcium ruthenate, a strontium ruthenium ternary compound, calcium iridate, strontium iridate, calcium platinate, strontium platinate, magnesium ruthenate, magnesium iridate, sodium ruthenate, sodium iridate, potassium iridate, potassium ruthenate, or combinations thereof. As illustrated in FIG. 1, the counter electrode 114 may comprise a monolithic body comprising a substantially uniform composition. In other embodiments, the counter electrode 114 may comprise a base material (e.g., a substrate) coated with counter electrode material formulated and configured to be substantially inert in the electrochemical cell 102. FIG. 2 is a simplified cross-sectional view of a counter electrode 114′, in accordance with some embodiments of the disclosure. The counter electrode 114′ may comprise a substrate (e.g., a core) 130 and a coating material 132. The coating material 132 may include one or more of the materials described above with reference to the counter electrode 114 of FIG. 1. For example, the coating material 132 may comprise ruthenium, osmium, rhodium, iridium, palladium, platinum, silver, gold, lithium iridate, lithium ruthenate, a lithium rhodate, a lithium tin oxygen compound, a lithium manganese compound, calcium ruthenate, a strontium ruthenium ternary compound, calcium iridate, strontium iridate, calcium platinate, strontium platinate, magnesium ruthenate, magnesium iridate, sodium ruthenate, sodium iridate, potassium iridate, potassium ruthenate, or combinations thereof. The coating material 132 may be substantially uniform over surfaces of the substrate 130. The coating material 132 may have a thickness T between about 1.0 mm and about 7.0 mm, such as between about 1.0 mm and about 3.0 mm, between about 3.0 mm and about 5.0 mm, or between about 5.0 mm and about 7.0 mm. In some embodiments, the thickness T is between about 3.0 mm and about 5.0 mm. The substrate 130 may comprise a material that is different than the coating material 132. The substrate 130 may comprise a metal or a non-metal. The substrate 130 may be selected to exhibit a coefficient of thermal expansion substantially similar to a coefficient of thermal expansion of the coating material 132. By way of nonlimiting example, the substrate 130 may comprise graphite (e.g., high density graphite), nickel, molybdenum, tantalum, chromium, tungsten, titanium, or another material. In some embodiments, the substrate 130 comprises high density graphite. Referring to FIG. 3, an electrochemical cell 300 for forming the counter electrode 114′ (FIG. 2) is illustrated. The electrochemical cell 300 includes an anode 302 and a cathode 304 immersed in a molten salt electrolyte 306. A power source 308, such as a direct current power source, may be operably coupled to the anode 302 and the cathode 304 for providing an electric potential between the anode 302 and the cathode 304 and a driving force for electrochemical reactions that may occur within the electrochemical cell 300. The electrochemical cell 300 may be configured as an electroplating electrochemical cell, wherein one or more components of the molten salt electrolyte 306 comprises a material to be deposited on the cathode 304. The molten salt electrolyte 306 may be formulated and configured to include at least one component of the coating material 132 (FIG. 2) of the counter electrode 114′ (FIG. 2). The molten salt electrolyte 306 may include, for example, a fluoride salt, chloride salt, a bromide salt, or another salt of one or more of ruthenium, osmium, rhodium, iridium, palladium, platinum, gold, and silver. By way of nonlimiting example, the molten salt electrolyte 306 may include ruthenium fluoride (RuF3, RuF4, RuF6, (RuF5)4), ruthenium chloride (RuCl2, RuCl3), ruthenium bromide (RuBr3), osmium fluoride (OsF6), osmium chloride (OsCl3, OsCl4), osmium bromide (OsBr3, OsBr4), rhodium chloride (RhCl3), rhodium (RhBr3), iridium fluoride (IrF3, IrF4, IrF6), iridium chloride (IrCl2, IrCl3, IrCl4), iridium bromide (IrBr3, IrBr4), palladium fluoride (PdF2, PdF4), palladium chloride (PdCl2), palladium bromide (PdBr2), platinum fluoride (PtF4, (PtF5)4, PtF6), platinum chloride (PtCl2, PtCl3, PtCl4), platinum bromide (PtBr2, PtBr3, PtBr4), silver fluoride (AgF), silver chloride (AgCl), silver bromide (AgBr), gold fluoride (AuF3, AuF5), gold chloride (AuCl, AuCl3), gold bromide (AuBr, (AuBr3)2), potassium hexachlorosmate (K2OsCl6), H2PtCl6, (NH4)2PtCl6, Pt(NO2)2(NH3)2, K2Pt(NO2)4, Na6Pt(SO3)4, Na6Pt(SO3)4, or combinations thereof. The anode 302 may include graphite, glassy carbon, nickel, cobalt, molybdenum, tantalum, tungsten, a platinum group metal, or another material. In some embodiments, the anode 302 comprises graphite. In some embodiments, where the anode 302 comprises molybdenum, tantalum, tungsten, or combinations thereof, the electrolyte 306 may comprise a bromide salt including at least one platinum group metal. In other embodiments, where the electrolyte 306 comprises dissolved platinum group metals dissolved in a chloride salt, a fluoride salt, or a combination thereof, the anode 302 may comprise graphite or glassy carbon. In further embodiments, where the anode 302 comprises a platinum group metal, the electrolyte 306 may comprise a bromide salt, a fluoride salt, a chloride salt, or combinations thereof. The cathode 304 may include any material on which the coating material 132 (FIG. 2) may be formed (i.e., a suitable substrate 130 (FIG. 2)). Stated another way, the cathode 304 may correspond to the substrate 130 (FIG. 2) and may include any of the materials described above with reference to the substrate 130. By way of nonlimiting example, the cathode 304 may include graphite (e.g., high density graphite), nickel, molybdenum, tantalum, titanium, chromium, tungsten, another material, or combinations thereof. Application of an electric potential between the anode 302 and the cathode 304 may cause dissolved ions (one or more of dissolved ruthenium, osmium, rhodium, iridium, palladium, platinum, silver, and gold ions) to move to the cathode 304, where such ions may be deposited to form a coating (e.g., the coating material 132 (FIG. 2)). After a sufficient duration, the cathode 304 may be coated to a desired thickness and the cathode 304 is removed from the electrochemical cell 300. In some embodiments, after removing the cathode 304 from the electrochemical cell 300, the cathode 304 is annealed, such as by exposing the cathode 304 to a temperature greater than about 1,000° C., greater than about 1,100° C., greater than about 1,200° C., greater than about 1,300° C., greater than about 1,400° C., greater than about 1,500° C., greater than about 1,600° C., greater than about 1,700° C., greater than about 1,800° C., greater than about 1,900° C., or even greater than about 2,000° C. In some embodiments, the annealing temperature may be less than a melting temperature of the coating material 132 (e.g., less than about 75% of the melting temperature of the coating material 132 in degrees Celsius). The cathode 304 may be exposed to the annealing temperature for a duration between about 2 hours and about 7 days, such as between about 2 hours and about 6 hours, between about 6 hours and about 12 hours, between about 12 hours and about 1 day, between about 1 day and about 2 days, between about 2 days and about 3 days, between about 3 days and about 5 days, or between about 5 days and about 7 days. The annealed cathode 304 may be used as a counter electrode (e.g., the counter electrode 114′ (FIG. 2)) in the electrochemical cell 102 (FIG. 1) for reducing metal oxides at the cathode 116 (FIG. 1). Although the counter electrode 114′ (FIG. 2) has been described as being formed in the electroplating electrochemical cell 300 (FIG. 3), the disclosure is not so limited. In other embodiments, the counter electrode 114′ may be formed by depositing the coating 132 (FIG. 2) on the substrate 130 (FIG. 1). By way of nonlimiting example, the coating 132 may be formed on the substrate 130 by chemical vapor deposition (CVD), physical vapor deposition (PVD), atomic layer deposition (ALD), plasma-enhanced chemical vapor deposition (PECVD), or another deposition method. FIG. 4 is a simplified flow diagram of a method 400 of reducing one or more metal oxides in an electrochemical cell, in accordance with embodiments of the disclosure. The method 400 includes act 402, including disposing one or more metal oxides in an electrochemical cell; act 404 including applying electric potential between a counter electrode and a working electrode of the electrochemical cell; and act 406 including reducing metal oxides at the cathode to form a substantially pure metal or a metal alloy. Act 402 may include disposing one or more metal oxides in an electrochemical cell. The electrochemical cell may be substantially similar to the electrochemical cell 102 described above with reference to FIG. 1. The metal oxide may comprise a metal oxide to be reduced, such as for example, uranium oxide, titanium monoxide, titanium dioxide, magnesium oxide, zirconium oxide, hafnium oxide, vanadium oxide, niobium oxide, tantalum pentoxide, chromium oxide, molybdenum oxide, tungsten oxide, ruthenium oxide, osmium oxide, rhodium oxide, iridium oxide, iron oxide, cobalt oxide, nickel oxide, silicon oxide, a lanthanide oxide, an actinide oxide, or combinations thereof. In some embodiments, the metal oxide comprises the working electrode of the electrochemical cell. In some such embodiments, the working electrode consists essentially of the metal oxide. In other embodiments, the metal oxide is directly secured to the working electrode, such as in a metal basket. In some such embodiments, the basket may comprise nickel, stainless steel, or another material. The electrochemical cell may include a molten salt electrolyte that may be substantially similar to those described above with reference to FIG. 1. For example, the molten salt electrolyte may comprise a mixture of lithium chloride and lithium oxide, a mixture of calcium chloride and calcium oxide, a eutectic mixture of sodium chloride and calcium chloride, a eutectic mixture of sodium chloride and potassium chloride, or combinations thereof. The counter electrode of the electrochemical cell may be substantially similar to the counter electrode materials described above with reference to FIG. 1. In some embodiments, the counter electrode comprises a platinum group metal. Act 404 may include applying an electric potential between the counter electrode and the working electrode of the electrochemical cell. The electric potential may be selected based on the composition of the molten salt electrolyte. By way of nonlimiting example, where the molten salt electrolyte comprises lithium chloride and lithium oxide, the electric potential (also referred to as the “cathodic potential”) may be between about 0.1 V and about 3.1 V, such as between about 0.1 V and about 1.5 V, between about 1.5 V and about 2.0 V, between about 2.0 V and about 2.5 V, between about 2.5 V and about 2.7 V, between about 2.7 V and about 2.9 V, or between about 2.9 V and about 3.1 V. Where the molten salt electrolyte comprises calcium chloride and calcium oxide, the electric (cathodic) potential may be between about 0.1 V and about 3.2 V, such as between about 0.1 V and about 2.0 V, between about 2.0 V and about 2.5 V, between about 2.5 V and about 2.7 V, between about 2.7 V and about 2.9 V, or between about 2.9 V and about 3.2 V. Where the molten salt electrolyte comprises eutectic mixture of sodium chloride and potassium chloride, the electric potential may be between about 0.1 V and about 3.2 V, such as between about 0.1 V and about 2.0 V, between about 2.0 V and about 2.5 V, between about 2.5 V and about 2.7 V, between about 2.7 V and about 2.9 V, or between about 2.9 V and about 3.2 V. Act 404 may further include maintaining a temperature of the electrochemical cell at or above a melting temperature of the molten salt electrolyte such that the molten salt electrolyte remains in a molten state. By way of nonlimiting example, the molten salt electrolyte may be maintained at a temperature between about 640° C. and about 680° C., such as between about 640° C. and about 660° C. or between about 660° C. and about 680° C. where the molten salt electrolyte comprises LiCl—Li2O. In some such embodiments, the molten salt electrolyte may be maintained at a temperature of about 650° C. In embodiments where the molten salt electrolyte comprises CaCl2—CaO, the molten salt electrolyte may be maintained at a temperature between about 800° C. and about 950° C., such as between about 800° C. and about 850° C., between about 850° C. and about 875° C., between about 875° C. and about 900° C., or between about 900° C. and about 950° C. Where the molten salt electrolyte comprises sodium chloride and calcium chloride, the molten salt electrolyte may be maintained at a temperature between about 550° C. and about 950° C., such as between about 550° C. and about 600° C., between about 600° C. and about 700° C., between about 700° C. and about 800° C., between about 800° C. and about 900° C., or between about 950° C. and about 950° C. Of course, it is contemplated that the temperature of the molten salt electrolyte may be maintained at a temperature other than the temperatures described above, depending on the composition of the molten salt electrolyte and desired operating conditions. Act 406 may include reducing metal oxides at the working electrode to form a substantially pure metal or metal alloy. In some embodiments, the metal may exhibit substantially the same composition as the metal oxide of the working electrode, except that the metal or metal alloy may be substantially free of oxygen. In some embodiments, the pressure within the electrochemical cell may be equal to about ambient pressure (i.e., atmospheric pressure). However, the disclosure is not so limited and the atmospheric pressure of the electrochemical cell may be different than those described above. Accordingly, in some embodiments, one or more metal oxides may be disposed on or proximate to the working electrode 116 (FIG. 1) of the electrochemical cell 102 (FIG. 1). The counter electrode 114 (FIG. 1) may be selected to comprise a material that is compatible with the molten salt electrolyte 110 (FIG. 1) and does not substantially corrode or wear (e.g., thin) responsive to exposure to the molten salt electrolyte 110. Without wishing to be bound by any particular theory, it is believed that forming the counter electrode 114 from the materials described herein facilitates operating the electrochemical cell 102 and reducing the metal oxides without substantially degrading or otherwise consuming the counter electrode 114. The counter electrode materials described herein may not substantially react with the molten salt electrolyte, oxide ions, or evolved reactive gases (e.g., selenium, tellurium, and iodine) at operating conditions of the electrochemical cell 102. Accordingly, such materials may maintain structural integrity during operation of the electrochemical cell 102. By way of comparison, conventional counter electrode materials (e.g., platinum, graphite, etc.) may react with at least one of one or more constituents of the molten salt electrolyte, oxide ions, or evolved gases, degrading the conventional counter electrode material. Advantageously, since the electrochemical cell 102 (FIG. 1) is configured as a direct oxide reduction cell, metal ions of the metal oxide are not substantially dissolved in the molten salt electrolyte 110 (FIG. 1). Accordingly, the metal oxide is reduced to the metal without metal ions dissolving in the molten salt electrolyte 110. Since the metal oxide is not dissolved in the molten salt electrolyte 110, in some embodiments, operation of the electrochemical cell 102 may consume less power than in other conventional electrochemical cells wherein the metal oxide being reduced is dissolved in the molten salt electrolyte. In addition, since the metal oxide is not dissolved, the size and shape of the metal formed may correspond to the size and shape of the metal oxide that is reduced in the electrochemical cell 102. Without wishing to be bound by any particular theory, it is believed that the counter electrode 114 materials described herein are substantially inert in the electrochemical cell 102 and are not substantially consumed because they do not react with or form complexes or compounds with the molten salt electrolyte 110. By way of nonlimiting example, iridium anode materials do not react in lithium chloride/lithium oxide molten salt electrolytes to form lithium iridate since the formation temperature of lithium iridate is higher than about 750° C. in such molten salts. Similarly, ruthenium counter electrode materials do not react with such molten salt electrolytes to form lithium ruthenate since the formation temperature of lithium ruthenate is higher than 650° C. Similarly, lithium rhodates, including dioxorhodates (e.g., Li2RhO2) and trioxorhodates (e.g., Li2RhO3), only form at a temperature greater than about 825° C. Accordingly, such materials can be used in the electrochemical cell 102 (FIG. 1) (e.g., an electrochemical cell with a LiCl—Li2O molten salt electrolyte having a temperature less than about 825° C.) since they do not form and react with the molten salt electrolyte. In addition, since lithium iridate, lithium ruthenate, and lithium rhodates are not formed at the operating conditions of the electrochemical cell, the counter electrode may comprise lithium iridate, lithium ruthenate, one or more lithium rhodates, or combinations thereof and may be substantially inert in the electrochemical cell. As another example, calcium ruthenate, strontium ruthenium ternary compounds, calcium iridate, strontium iridate, calcium platinate, and strontium platinate form at high gas pressures (such as at pressures between about 1.0 GPa and about 5.0 GPa) and do not substantially form at operating conditions of an electrochemical cell with either a calcium chloride-calcium oxide or a sodium chloride-calcium chloride molten salt electrolyte. Accordingly, counter electrodes comprising one or more of the materials described above in such molten salt electrolytes may be substantially inert under operating conditions of the electrochemical cell. In addition, counter electrode materials comprising calcium ruthenate, strontium ruthenium ternary compounds, calcium iridate, strontium iridate, calcium platinate, and strontium platinate may be substantially inert in the electrochemical cell, since they form at gas pressures above operating pressures of the electrochemical cell 102 and do not react with the molten salt electrolyte 110 at operating conditions of the electrochemical cell 102. An electrolyte comprising 200 g of high purity anhydrous calcium chloride was prepared. Calcium oxide was added to the electrolyte such that the calcium oxide constituted about 1.0 weight percent of the electrolyte. The electrolyte was melted in an alumina/nickel crucible inside a glove box under an argon atmosphere having less than 0.1 ppm moisture and oxygen. A first monolithic counter electrode (anode) comprising a monolithic rod of iridium and a second counter electrode comprising a monolithic rod of ruthenium, a working electrode (cathode) comprising tantalum pentoxide pellets in a basket, and a reference electrode comprising glassy carbon were cleaned in an ultrasonic bath, oven dried, and disposed in the electrochemical cell. The working electrode included about 18.1 weight percent oxygen. The first counter electrode, the second counter electrode, and reference electrode rods had a diameter of about 3.0 mm and a length of about 100 mm. The working electrode was prepared by sintering oxides of tantalum in air or in a reducing atmosphere to form sintered pellets. The sintered pellets were cylindrical in shape and had a diameter between about 13.0 mm and about 15.0 mm and a length between about 1.0 mm and about 5.0 mm. The first counter electrode, the second counter electrode, the working electrode, and reference electrode were sheathed in high purity alumina tubes. The electrolyte was heated in a furnace to melt the electrolyte and form a molten salt electrolyte. The molten salt electrolyte was maintained at a temperature between about 800° C. and about 950° C. The first counter electrode, the second counter electrode, the working electrode, and the reference electrode were disposed in the electrochemical cell such that the lower portions of the respective electrodes were in contact with the molten salt electrolyte. The cell voltage was controlled between about 2.5 V and about 3.0 V. The current was measured as a function of time, as illustrated in the graph of FIG. 5. The electrochemical cell was operated inside an argon atmosphere in a glove box. As illustrated in FIG. 5, the initial current measured exhibited a sharp rise before declining over time. The initial peak in the current is due to the non-conducting nature of the metal oxide and the reduction in current correlates to the reduction of the metal oxide to a conducting metal. The residual oxygen content of the metal oxide was measured with a LECO analyzer. The metal oxide was reduced by about 98.9%. In other words, after the electrochemical reaction, the metal oxide had less than about 2,000 ppm oxygen, indicating a reduction in oxygen content of about 98.9%. Stated another way, the metal oxide was reduced and included about 0.2 weight percent oxygen. The first counter electrode and second counter electrode were visually inspected to record the occurrence of any possible mechanical degradation such as cracking, thinning, corrosion, erosion, or necking. No mechanical degradation was observed in the anodes. A first counter electrode (anode) comprising platinum was disposed in an electrochemical cell including a molten salt electrolyte comprising lithium chloride and lithium oxide. The electrochemical cell included depleted uranium oxide disposed in a working electrode (cathode) basket which was in contact with the cathode lead. A second iridium counter electrode was disposed in the same electrochemical cell. The electrochemical cell was in an argon atmosphere in a glove box. A current was applied between and the single cathode and each of the platinum counter electrode and the iridium counter electrode and the metal oxide was reduced in the electrochemical cell. After the metal oxide was substantially reduced, both the platinum counter electrode and the iridium counter electrode were removed and analyzed for mechanical degradation and thinning. The platinum counter electrode exhibited a reduction in diameter of about 17 percent. The iridium counter electrode exhibited a reduction in diameter of less than about 1 percent. Accordingly, the iridium counter electrode exhibited substantially less thinning than the platinum counter electrode. A first counter electrode (anode) comprising monolithic iridium and a second counter electrode (anode) comprising monolithic ruthenium were disposed in an electrochemical cell including a molten salt electrolyte comprising lithium chloride and lithium oxide. The first counter electrode and the second counter electrode comprised rods with a diameter of 3.0 mm and a length of about 100 mm. The molten salt electrolyte comprised about 1.0 weight percent lithium oxide. The temperature of the molten salt electrolyte was about 650° C. The electrochemical cell included about 100 grams of unirradiated depleted uranium oxide (UO2) disposed in a stainless steel basket which was in contact with the cathode lead. The electrochemical cell was in an argon atmosphere in a glove box. A current was applied between the single cathode and each of the first counter electrode (monolithic iridium) and the second counter electrode (monolithic ruthenium) and the uranium oxide was reduced in the electrochemical cell. After about 80 hours, both the first counter electrode and the second counter electrode were removed and analyzed for mechanical degradation and thinning. No perceptible mechanical degradation was observed in either of the first counter electrode or the second counter electrode. Accordingly, the first counter electrode and the second counter electrode exhibited substantially less thinning than a platinum electrode (such as the platinum electrode of Example 2). An electrochemical cell comprising a molten salt electrolyte including a mixture of LiCl and about 1.0 weight percent Li2O, about 0.1 weight percent Na2Se, about 0.1 weight percent Li2Te, and about 0.1 weight percent LiI was prepared. The working electrode (cathode) of the electrochemical cell included pellets and/or chunks of titanium monoxide, manganese oxide, and nickel oxide (NiO) packed into a stainless steel basket. The stainless steel basket was in electrical communication with the working electrode through a stainless steel wire used as the working electrode current collector. Three counter electrode (anodes) were disposed in the electrochemical cell: one counter electrode comprising monolithic iridium; one counter electrode comprising monolithic ruthenium; and one counter electrode comprising monolithic platinum. The counter electrodes each had a diameter of about 3.0 mm and a length of about 100.0 mm. The electrochemical cell was operated to reduce the oxides in the stainless steel basket. Selenium, tellurium, and iodine gases evolved during operation of the electrochemical cell. After the metal oxides in the stainless steel basket were reduced, the counter electrodes were inspected. The platinum counter electrode exhibited significant thinning (e.g., necking), while the ruthenium counter electrode and the iridium counter electrode did not exhibit any detectable material loss. While embodiments of the disclosure may be susceptible to various modifications and alternative forms, specific embodiments have been shown by way of example in the drawings and have been described in detail herein. However, it should be understood that the disclosure is not limited to the particular forms disclosed. Rather, the disclosure encompasses all modifications, variations, combinations, and alternatives falling within the scope of the disclosure as defined by the following appended claims and their legal equivalents.
	description	The present application is a continuation-in-part of U.S. application Ser. No. 15/282,814, titled NEUTRON REFLECTOR ASSEMBLY FOR DYNAMIC SPECTRUM SHIFTING, filed Sep. 30, 2016. U.S. application Ser. No. 15/282,814 claims the benefit of U.S. Provisional Patent Application No. 62/337,235, titled “NEUTRON REFLECTOR ASSEMBLY FOR DYNAMIC SPECTRUM SHIFTING”, filed May 5, 2016; and U.S. Provisional Patent Application No. 62/234,889, entitled “MOLTEN CHLORIDE FAST REACTOR AND FUEL” and filed on Sep. 30, 2015. In addition, the present application claims the benefit of U.S. Provisional Patent Application No. 62/330,726, titled “IMPROVED MOLTEN FUEL REACTOR CONFIGURATIONS”, filed May 2, 2016. A particular classification of fast nuclear reactor, referred to as a “breed-and-burn” fast reactor, includes a nuclear reactor capable of generating more fissile nuclear fuel than it consumes. That is, the neutron economy is high enough to breed more fissile nuclear fuel (e.g., plutonium-239) from fertile nuclear reactor fuel (e.g., uranium-238) than it burns in a fission reaction. In principle, a breed-and-burn reactor may approach an energy extraction rate of 100% of the fertile materials. To initiate the breeding process, a breed-and-burn reactor must first be fed with an amount of fissile fuel, such as enriched uranium. Thereafter, breed-and-burn reactors may be able to sustain energy production over a timespan of decades without requiring refueling and without the attendant proliferation risks of conventional nuclear reactors. One type of breed-and-burn reactor is a molten salt reactor (MSR). Molten salt reactors are a class of fast spectrum nuclear fission reactors wherein the fuel is a molten salt fluid containing mixed or dissolved nuclear fuel, such as uranium or other fissionable elements. In an MSR system, the unmoderated, fast neutron spectrum provided by fuel salts enables good breed performance using the uranium-plutonium fuel cycle. In contrast to the fast spectrum neutrons that dominate breeding of fissile fuel from fertile fuel, thermal neutrons dominate the fission reaction of fissile fuel. A fission reaction resulting from a collision of a thermal neutron with a nuclide can consume the fissile fuel in a fission reaction, releasing fast spectrum neutrons, gamma rays, large amounts of heat energy and expelling fission products, such as smaller nuclei elements. Consuming nuclear fuel is referred to as burnup or fuel utilization. Higher burnup typically reduces the amount of nuclear waste remaining after the nuclear fission reaction terminates. The fast neutron spectrum also mitigates fission product poisoning to provide exceptional performance without online reprocessing and the attendant proliferation risks. The design and operating parameters (e.g., compact design, low pressures, high temperatures, high power density) of a breed-and-burn MSR, therefore, offer the potential for a cost-effective, globally-scalable solution to zero carbon energy. During operation of an MSR system, molten fuel salt exchange can allow some control over reactivity and breeding in the reactor core within desired operational bounds by altering the composition of the circulating molten fuel salt. In some implementations, the reactor core is wholly or partially enclosed in a neutron reflector assembly containing a neutron reflector material. The disclosed dynamic neutron reflector assembly allows additional dynamic and/or incremental control over reactivity and breed rate by adjusting reflectivity characteristics of a neutron reflector assembly to manage the neutron spectrum in the reactor core. Such control manages the reactivity and the breed rate in the reactor core. The composition of materials in the dynamic neutron reflector assembly may be altered by selectively inserting or removing neutron-spectrum-influencing materials, such as neutron reflectors, moderators or absorbers, to dynamically manage the dynamic neutron reflector assembly's neutron-spectrum-influencing characteristics (“reflectivity characteristics”). Alternatively, these reflectivity characteristics may be adjusted by varying the temperature, density, or volume of the material in the dynamic neutron reflector assembly. In some implementations, the dynamic neutron reflector assembly may include a flowing neutron reflector material that is in thermal contact with the fuel (e.g., molten fuel salt). The flowing neutron reflector material may be in any appropriate form including, without limitation, fluids like lead bismuth, slurry of suspended particulates, solids such as a powder, and/or pebbles such as carbon pebbles. The dynamic neutron reflector assembly may selectively circulate or flow through the assembly one or more neutron absorbing materials, such that it is possible to selectively add or remove reflector material therefrom. In other implementations, the flowing neutron reflector material can extract heat from the molten fuel salt in a heat exchanger via a primary or secondary coolant circuit. FIG. 1 is a schematic view of an example molten salt reactor (MSR) system 100 enabling an open breed-and-burn fuel cycle with fuel feed 102 and fuel outlet 104. The fuel outlet 104 flows molten fuel salt 108 from a reactor vessel 107 through a primary coolant loop to an external heat exchanger (not shown), which extracts heat (e.g., for use in a steam turbine) and cools the molten fuel salt 108 for return to the reactor vessel 107 via the fuel feed 102. The molten fuel salt 108 flows into the reactor vessel 107 through a molten fuel salt input 111 and flows out of the reactor vessel 107 through a molten fuel salt output 113. The reactor core section 106 is enclosed by the reactor vessel 107, which may be formed from any material suitable for use in molten salt nuclear reactors. For example, the bulk portion of the reactor core section 106 may be formed from one or more molybdenum alloys, one or more zirconium alloys (e.g., Zircaloy), one or more niobium alloys, one or more nickel alloys (e.g., Hastelloy N) or high temperature steel and other similar materials. The internal surface 109 of the reactor core section 106 may be coated, plated or lined with one or more additional material in order to provide resistance to corrosion and/or radiation damage. Reactor core section 106 is designed to maintain a flow of a molten fuel salt 108, wherein such flow is indicated by hollow tip thin arrows as in FIG. 1. In one implementation, the reactor vessel 107 enclosing the reactor core section 106 may have a circular cross-section when cut along a vertical or Z-axis (i.e., yielding a circular cross-section in the XY plane), although other cross-sectional shapes are contemplated including without limitation ellipsoidal cross-sections and polygonal cross-sections. As part of the reactor startup operation, the MSR system 100 is loaded with an enriched fuel charge of initial molten fuel, such as uranium-233, uranium-235, or plutonium-239. In one implementation, uranium-235 is used as a startup fuel in the form of PuCl3, UCl4, UCl3, and/or UF6 along with a carrier salt (e.g., NaCl, NaF, etc.). In one example, the initial molten fuel mixture contains enriched uranium at 12.5 w %, although other compositions may be employed. The initial molten fuel circulates through the reactor core section 106 of the MSR system 100, ignited by the criticality or reactivity of thermal neutrons of the enriched uranium. During operation, the initial molten fuel may be augmented by the breed-and-burn processes and by extraction and supplementation of molten fuel salt in varying compositions and amounts, in one approach, to managing the reactivity in the reactor core section 106. A neutron reflector assembly 110 is disposed at or near the exterior of the reactor core section 106, such that the neutron reflector assembly 110 surrounds at least a portion of the nuclear fission region within the reactor core section 106. The neutron reflector assembly 110 may be designed in a single contiguous piece or may be composed of multiple segmented reflectors as explained in more detail below. The neutron reflector assembly 110 may be formed from and/or include any material suitable for neutron reflection, neutron moderation and/or neutron absorption, such as, for example, one or more of zirconium, steel, iron, graphite, beryllium, tungsten carbide, lead, lead-bismuth, etc. Among other characteristics, the neutron reflector assembly 110 is suitable for reflecting neutrons emanating from the reactor core section 106 back into the molten fuel salt 108, according to dynamic incrementally adjustable reflectivity characteristics. One type of a dynamic incrementally adjustable reflection characteristic is neutron reflection, an elastic scattering of neutrons as they collide with reflector nuclei. Colliding neutrons are scattered at substantially the same energy with which they arrived but in a different direction. In this manner, a high percentage of fast spectrum neutrons can be reflected back into the reactor core section 106 as fast spectrum neutrons, where they can collide with fertile nuclear material to breed new fissile nuclear material. Accordingly, neutron reflector material in the neutron reflector assembly 110 can enhance the breed operation of a breed-and-burn fast reactor. Additionally, or alternatively, another dynamically adjustable reflection characteristic is neutron moderation, an inelastic scattering of neutrons as they collide with moderator nuclei. Colliding neutrons are scattered at a lower energy than that with which they arrived (e.g., a fast spectrum neutron scatters as a thermal spectrum neutron) and with a different direction. In this manner, a high percentage of fast spectrum neutrons can be reflected back into the reactor core section 106 as thermal neutrons, where they can collide with fissile nuclear material and result in a fission reaction. Accordingly, neutron moderator material in the neutron reflector assembly 110 can enhance the burn-up operation of a breed-and-burn fast reactor. Additionally or alternatively, another dynamically adjustable reflection characteristic is neutron absorption, also known as neutron capture: a nuclear reaction in which an atomic nucleus and one or more neutrons collide and merge to form a heavier nucleus. Absorbed neutrons are not scattered but remain part of the merged nuclei unless released at a later time, such as part of a beta particle. Neutron absorption provides the reflectivity characteristic of zero or minimal reflection. In this manner, fast and thermal neutrons emanating from the reactor core may be prevented from scattering back into the reactor core section 106 to collide with fissile or fertile material. Accordingly, neutron absorbing material in the neutron reflector assembly 110 can diminish the breed operation and burn operation of a breed-and-burn fast reactor. Dynamic control over neutron reflectivity characteristics of the neutron reflection assembly 110 permits selection of a desired reactivity level in reactor core section 106. For example, molten fuel salt 108 requires a minimum level of thermal neutron contact to remain critical in reactor core section 106. The dynamic neutron reflector assembly 110 may be adjusted to provide the reflectivity characteristics for maintaining or contributing to the criticality in the molten fuel salt 108 within the reactor core section 106. As another example, it may be desired to operate the MSR system 100 at full power, which would motivate an increased thermalization of neutrons in the reactor core section 106 to increase the fission rate. The reflectivity characteristics of dynamic neutron reflector assembly 110 could be therefore increased to provide more moderation until a desired reactivity level representing full power for the reactor core section 106 has been reached. In contrast, as MSR system 100 is a breed-and-burn reactor, it may be desired to dynamically control breed rate at various points over the lifecycle of the reactor. For example, early in the reactor's lifecycle, a high breed rate may be desired to increase the availability of fissile material in reactor core section 106. The reflectivity characteristics of dynamic neutron reflector 110 may therefore be adjusted to provide increased reflection of fast neutrons into reactor core section 106 to breed more fertile material into fissile fuel. As more fast neutrons are reflected into reactor core section 106 over time, the fast neutrons may breed fertile material into fissile material until a desired concentration of fissile material has been reached. Later in the reactor's lifecycle, it may be desirable to increase burnup to provide increased power through increased burnup. The reflectivity characteristics of dynamic neutron reflector assembly 110 may therefore be adjusted to increase moderation of fast neutrons into thermal neutrons to maintain the desired burn rate. In this way, the core reactivity and the ratio of breeding to burning may be accurately controlled over time by adjusting the reflectivity characteristics of dynamic neutron reflector assembly 110. For example, an operator of the MSR system 100 may wish to maintain a high and consistent burn profile over time. In some implementations, a desired burn profile is a burn profile that remains near maximum burn rate of the MSR system 100 over an extended period of time, such as over a period of years or decades. Reflectivity characteristics of dynamic neutron reflector assembly 110 may be chosen at various intervals over the extended period of time to obtain such a burn profile. As in the example above, early in the life cycle of the MSR system 100, reflectivity characteristics may be chosen to reflect more fast neutrons into reactor core section 106 to breed fertile material into fissile material until a desired concentration of fissile material has been reached. Reflectivity characteristics may be again adjusted for increased thermalization appropriate to the concentration of fissile material. Over time, as the fissile material is burned, reflectivity characteristics of dynamic neutron reflector assembly 110 may again be adjusted to introduce more breeding through fast neutron reflection, by reducing moderation and/or increasing fast neutron reflection. These adjustments may continue such that the burn profile of MSR system 100 remains high, and fertile material is bred into fissile material at a rate sufficient to supply the MSR system 100 with fuel over the extended period. FIG. 2 is a plot 200 of reactivity against time of a fast spectrum MSR with one or more dynamic reflector assemblies against two other assembly configurations with static neutron influencing characteristics. A plot line 202 shows reactivity over time for a fast spectrum MSR reactor with static lead neutron reflector assembly surrounding a reactor core, wherein the lead neutron reflector assembly tends to elastically scatter fast neutrons into the reactor core. After a time T0, when the reactor is started with an initial fuel charge, breeding of fertile fuel may occur rapidly due to reflection of fast neutrons into the reactor core. After T1, reactivity on the plot line 202 gradually increases over time as the breeding increases the amount of available fissile material to burn, reaching a maximum at a time near T4. Breeding may slow over time with increasing burnup as fertile fuel previously present in the reactor core is converted to fissile material or fissioned due to increased competition for neutrons with products of fission. The plot line 202 does not show a constant reactivity level over time because, near the beginning of the period, there are not sufficient fast neutrons in the fuel region to breed enough fissile material to support a high burn rate. Over time, the larger number of fast neutrons breeds fertile material into fissile material, and reactivity increases but remains below the maximum burn rate of which the reactor is capable. Near the end of the period, around time T5, reactivity reaches a local maximum and begins to decline as the supply of fertile material begins to decline. A plot line 204 shows reactivity over time for a fast MSR with a static graphite moderator configuration, wherein the moderating neutron reflector assembly tends to provision the reactor core with thermalized neutrons. On the plot line 204, reactivity begins around time T0 at a relatively higher level than plot line 202 due in part to thermalization caused by the graphite moderator increasing the probability of fission. Plot line 204 may drop significantly near time T0 due to thermal spectrum multiplication adjacent to the graphite reflector. Reactivity may then gradually reduce over time in a generally linear manner as the thermal neutrons burn fissile fuel in the reactor core. The plot line 204 is similar to plot line 202 in the respect that neither plot line reaches or maintains a maximized burn rate achievable within the reactor core. The plot line 204 does not reach the reactor's maximum burn rate because there are not enough fast neutrons to maintain a breeding rate high enough to support the burn rate as time progresses though the period T0-T5. In the plot lines 202 and 204, the burn rate is not optimized over the time period T0-T5. Instead, each plot has a period of relatively higher burn rate and a period of relatively lower burn rate over the course of the graph. The plot lines 202 and 204 are shown in contrast to plot line 206. The plot line 206 illustrates reactivity over time for a fast MSR system with a dynamic neutron reflector assembly, starting with a high moderator configuration and changing to a high reflector configuration, thereafter being dynamically controlled to achieve desired reactivity conditions within the reactor core. Reactivity over time on the plot line 206 starts relatively high after an initial fuel charge is loaded around time T0, and remains high due to the dynamically controllable nature of the reflection and thermalization of neutrons. Around time T0, the composition of material in the neutron reflection assembly is adjusted for a moderation rate that correlates with the concentration of fissile material available in the fuel region at that time. As the burn up progresses, the composition of material in the neutron reflection assembly is adjusted to increase fast neutron reflection and decrease moderation to continue supplying the fuel region with newly bred fissile material while, at the same time, maintaining an appropriate amount of thermalization to match the current conditions in the fuel region. The adjustments may be performed continuously or as a batch process, and continue over time towards T5. An effect of these dynamic neutron reflector assembly adjustments is to maintain a relatively stable and high reactivity rate over the entire period T0-T5 that is not feasible with static moderators and neutron reflectors, such as those represented by the plot lines 202 and 204, respectively. Nevertheless, the same dynamic neutron reflector assembly may be used to control reactivity in other ways (e.g., to reduce reactivity, etc.). It should also be noted that inclusion of a neutron absorber within the neutron reflector assembly can also impact the reactivity within the reactor core. Dynamic adjustments among neutron reflector, moderator, and absorber materials within the neutron reflector assembly can provide richer control options than static neutron reflector assemblies alone. FIG. 3 is a schematic view of a segmented dynamic neutron reflector assembly 300 surrounding an MSR core 301. The MSR core 301 is equipped with a fuel feed 308 and a fuel outlet 310. The fuel outlet 310 flows molten fuel salt from a reactor vessel 303 through a primary coolant loop to an external heat exchanger (not shown), which extracts heat (e.g., for use in a steam turbine) and cools the molten fuel salt for return to the reactor vessel 303 via the fuel feed 308. The molten fuel salt flows into the reactor vessel 303 through a molten fuel salt input 312 and flows out of the reactor vessel 303 through a molten fuel salt output 314. Segmented dynamic neutron reflector 300 may partially or substantially surround the MSR core 301. For example, there may be gaps between the segments 302, 304, 306 or the segments 302, 304, 306 may encircle the MSR core contiguously. Although three segments of the dynamic reflector assembly 300 are shown in FIG. 3, it should be understood that the dynamic reflector assembly may comprise any number of segments. The segments of the dynamic reflector assembly 300 may surround the core by completely or partially encircling the core radially. Segments of the dynamic reflector assembly 300 may be optionally positioned above and/or below the reactor core in combination with, or instead of, radial reflector segments. It should be understood that in some cases it may not be possible for the segmented dynamic neutron reflector to completely surround the reactor core in an uninterrupted or completely contiguous manner. For example, it may be appropriate to dispose various structures and instruments around the fast MSR core 301 with supporting elements such as input/output piping, power supply conduits, data conduits, and/or other instrumentation, controls, and supporting hardware. These structures and instruments may require direct or indirect access to the reactor core such that the segments of the dynamic reflector assembly 300 may need to be shaped or positioned to accommodate access. Accordingly, in some implementations, it may be appropriate to permit gaps between the segments or arrangements wherein portions of the area surrounding the reactor core are not covered by segments of the dynamic reflector assembly 300. Some or each segment 302, 304, 306 of the dynamic reflector assembly 300 may contain one or more channels (not shown in FIG. 3) for conducting a flowing reflector material. As used in this application, the term channels refers not only to a tubular enclosed passage, but to any volume suitable for flowing a reflector material. A flowing reflector material may include materials that may not necessarily be fluids, but materials that can circulate or flow through the assembly, such that it is possible to selectively add or remove reflector material therefrom. Examples of suitable neutron reflector materials include fluids, slurry of suspended particulates, and/or solids such as a powder, and/or pebbles, such as carbon pebbles, etc. The segments 302, 304, 306 may contain one or more first channels for conducting a flowing reflector material in a first direction, such as, for example, down along the periphery of the respective segments, and one or more second channels for conducting a flowing reflector material in a second direction, such as, for example, back up to the top of dynamic neutron reflector assembly 300. The channels of the various reflector segments may be fluidically coupled such that the flowing neutron reflector material flows between the segments. In another implementation, the reflector segments may be fluidically separate from one another such that flowing reflector material flows into and out of only a single segment. In an implementation, one or more of the fluid channels in the reflector segments may be in thermal communication with a heat exchanger and/or the molten fuel salt, acting as a coolant. The flowing reflector material may thus exchange heat with the molten fuel salt, and transfer the heat via the heat exchangers to a secondary coolant circuit to supply heat from the reactor to a turbine or other electricity generating equipment. As the flowing reflector material exchanges heat with the reactor core through a primary and/or a secondary coolant circuit, the flowing reflector material temperature may fluctuate. As the flowing reflector material's temperature fluctuates, its density may vary. For example, in an implementation, the flowing reflector material is molten lead-bismuth, and the molten lead-bismuth will experience a higher density at lower temperatures. As the temperature of the molten lead-bismuth lowers and its density rises, the number of molecules per unit volume of the lead-bismuth will increase. As the number of molecules per unit volume increases (i.e., higher density), the likelihood of reflecting a fast spectrum neutron emanating from the reactor core increases, thus increasing the effective reflectivity of the flowing reflector material without changing the volume of the material. In another implementation, the density of the flowing reflector material may be adjusted by introducing a non-reflective material (such as non-reflective material particulates, fluids gas bubbles, etc.) into the flowing reflector material. In yet another implementation, the density of the flowing reflector material may be adjusted by adjusting environmental characteristics to vaporize the flowing reflector material into a low density vapor phase. In this way, the material composition of the dynamic neutron reflector assemblies, and thus its reflectivity, may be altered. FIG. 4 illustrates an MSR system 400 with a dynamic flowing neutron reflector assembly 406 equipped with a spillover reservoir 408. A molten fuel salt 402 flows in an upward direction as it is heated by the fission reaction in the internal central reactor core section and flows downward as it cools around the internal periphery of the reactor vessel 401. In FIG. 4, hollow tip arrows indicate the flow of molten fuel salt through MSR system 400. The constituent components of the molten fuel may be well-mixed by the fast fuel circulation flow (e.g., one full circulation loop per second). In one implementation, one or more heat exchangers 404 are positioned at the internal periphery of the reactor vessel 401 to extract heat from the molten fuel flow, further cooling the downward flow, although heat exchangers may additionally or alternatively be positioned outside the reactor vessel 401. MSR system 400 includes dynamic neutron reflector assemblies 406. Operating temperatures of MSR system 400 may be high enough to liquefy a variety of suitable neutron reflector materials. For example, lead and lead-bismuth melt at approximately 327° C. and 200° C., respectively, temperatures within the operating range of the reactor. In an implementation, dynamic neutron reflector assemblies 406 are configured to contain a flowing and/or circulating fluid-phase of the selected neutron reflector materials (e.g., lead, lead-bismuth, etc.). In FIG. 4, solid tip arrows indicate the flow of neutron reflector material. Dynamic neutron reflector assemblies 406 may be formed from any suitable temperature and radiation resistant material, such as from one or more refractory alloys, including without limitation one or more nickel alloys, molybdenum alloys (e.g., a TZM alloy), tungsten alloys, tantalum alloys, niobium alloys, rhenium alloys, silicon carbide, or one or more other carbides. In an implementation, dynamic neutron reflector assemblies 406 are positioned on, and distributed across, the external surface of the reactor core section. In implementations, the dynamic neutron reflector assemblies 406 may be segmented, as explained above with reference to FIG. 3. In an implementation, dynamic neutron reflector assemblies 406 are arranged radially across the external surface of the reactor core section. Dynamic neutron reflector assemblies 406 may be arranged to form a contiguous volume of neutron reflector material around the reactor core section. Any geometrical arrangement and number of dynamic neutron reflector assemblies 406 is suitable for the technology described herein. For example, dynamic neutron reflector assemblies 406 may be arranged in a stacked ring configuration, with each module filled with a flow of neutron reflector material to form a cylindrical neutron reflecting volume around the reactor core section. Dynamic neutron reflector assemblies 406 may also be arranged above and below the reactor core section. The composition of the dynamic neutron reflector assemblies 406 may be adjusted to change reflectivity characteristics, such as, for example, by adjusting the volume of the flowing reflector material in reflectors 406. One way of adjusting the volume of the flowing reflector material in reflectors 406 is to pump the material into or out of dynamic reflectors 406 into spillover reservoir 408 via piping assembly 410 and pump 414. To decrease the volume of the flowing neutron reflector material, and thus to decrease the reflectivity characteristics of reflectors 406, a portion of the flowing neutron reflector material may be pumped or displaced into spillover reservoir 408 via piping assembly 410. A valve 412 and pump 414 may cooperate to regulate the flow of the flowing neutron reflector material through piping assembly 410. To increase the volume of the flowing neutron reflector material, valve 412 and pump 414 may cooperate to flow the flowing neutron reflector material out of overflow tank 408 and back into reflectors 406 via piping assembly 410. In another implementation, the reflectivity of dynamic neutron reflector assemblies 406 may be adjusted by regulating the temperature, and thus the density, of the flowing neutron reflector material. Changes in the density of the flowing neutron reflector material alter its neutron reflective characteristics as denser materials have a higher mass per unit volume. Denser materials will contain more molecules per unit volume, and are therefore more likely to reflect neutrons because any neutron travelling through the denser material will be more likely to strike a molecule of the flowing neutron reflector material and thus be reflected. Pump 414 and valve 412 may cooperate to increase or decrease the flow rate of the flowing neutron reflector material into or out of dynamic neutron reflectors 406 to regulate the temperature of the reflecting flowing neutron reflector material. In other implementations, spillover reservoir 408 may be replaced with other configurations, such as a closed circuit loop. The MSR system 400 may include a flowing neutron reflector material cleaning assembly 416. The flowing neutron reflector material cleaning assembly 416 is in fluid communication with the piping assembly 410, and may be located on either side of valve 412 and pump 414. The flowing neutron reflector material cleaning assembly 416 may filter and/or control the chemistry of the neutron reflector material. For example, the flowing neutron reflector cleaning assembly 416 may remove oxygen, nitrites, and other impurities from the neutron reflector material. In an implementation, a zircon nitrite coating in the neutron reflector cleaning assembly 416 is configured to control the chemistry of the flowing neutron reflector material. In another implementation, the flowing neutron reflector cleaning assembly 416 may perform a “slagging” technique wherein the flowing neutron reflector cleaning assembly 416 captures oxygen as an oxide material. If the oxide material is molten, it may phase separate and the flowing neutron reflector cleaning assembly 416 may remove the oxide material from the neutron reflector material by, for example, scraping the oxide material. In another implementation, the flowing neutron reflector cleaning assembly 416 is configured for a hydrogen treatment of the neutron reflector material to remove oxygen contained therein. The composition of dynamic neutron reflectors 406 may also be adjusted by introducing a flowing moderator material. The flowing moderator material may be held in a reserve tank (not shown) and introduced into dynamic neutron reflectors 406 via piping assembly 410 and pump 414 in fluid communication with the fluid moderator reserve tank. The flowing moderator material may circulate in dynamic reflectors 406, and may be removed by pump 414 into the reserve tank via piping assembly 410. In an implementation, water or heavy water may be used as a flowing moderating liquid in dynamic neutron reflectors 406. In another implementation, beryllium may be used as a flowing moderating material in dynamic neutron reflectors 406. In yet another implementation, LiF—BeF2 may be used a flowing moderating material in dynamic neutron reflectors 406 and/or in the fuel salt itself. The pump 414 may pump the flowing moderator liquid and/or the flowing neutron reflector material into and out of the dynamic reflectors 406 continuously and/or in a batch process. As previously described, neutron absorbing material can also be incorporated into dynamic neutron reflector assemblies 406, individually or in combination with various compositions and/or configurations of neutron reflector materials and neutron moderator materials. FIG. 5 is a top-down schematic view of a dynamic neutron reflector assembly 500 with a plurality of refractory clad sleeves 502 to conduct a flowing neutron reflector material there through. In an implementation, flowing neutron reflector assembly 500 substantially surrounds a nuclear fuel region 504 from which fast spectrum neutrons 506 emanate. In FIG. 5, example paths of fast spectrum neutrons 506 are indicated by lines terminating in double arrows, such as lines 508. The example fast spectrum neutrons 506 are inelastically scattered (or reflected) from the flowing reflector material and back into the nuclear fuel region 504. The reflector configuration of FIG. 5 may be used to incrementally shift neutron spectrum in nuclear fuel region 504 by selectively filling each of the channels 502 with a volume of neutron reflector material. In FIG. 5, the neutron reflector material flows upward through a refractory clad channel 502 toward the viewer. In an implementation, neutron reflector material may circulate in channels 502 (e.g., cells, sleeves, conduits, etc.) with input and output ports above the nuclear fuel region 504 such that no fixtures or ports are needed beneath the reactor. In other implementations, the neutron reflector material may flow in only one direction, either in an upward or downward direction, through the channels 502 with one port above the nuclear fuel region 504 and another port below fuel region 504. In yet other implementations, the neutron reflector material may comprise a semi-stagnant or creeping flow through the channels 502. In yet other implementations, the neutron reflector material may flow through radial input and output ports. The dynamic neutron reflector assembly 500 is in thermal communication with heat exchanger 510 disposed on the opposite side from fuel region 504. The heat exchanger 510 may contain one or more types of liquid coolant circulating there through. As neutron reflector 500 exchanges heat with the heat exchanger 510, the heat exchanger 510 may transport the heat away from the dynamic neutron reflector assembly 500 as part of a secondary coolant circuit. The secondary coolant circuit may supply heat to electricity generation equipment, such as, for example, a steam-driven turbine. In an implementation, molten fuel salt may flow upward through the nuclear fuel region 504 and downward through the heat exchanger 510, thus exchanging heat as part of a primary coolant circuit. In other words, the heat exchangers may exchange heat with both the molten fuel salt and exchange heat with the flowing neutron reflector in the channels 502. The flow rate of neutron reflector material may be adjusted to vary contact time with the heat exchangers to vary the temperature of reflector material flowing in the channels 502. As the temperature of reflector material varies, its density changes accordingly. Changes in the density of the reflector material alter its neutron reflective characteristics as denser materials have a higher mass per unit volume and are therefore more likely to reflect neutrons. The channels 502 may be formed in geometric shapes including without limitation square, rectangular, round, circular, polygonal, etc. FIG. 6 is a top-down schematic view of a dynamic neutron reflector assembly 600 with a plurality of sleeves 602 conducting a flowing neutron reflector material and a plurality of sleeves 604 including neutron moderating members 606 selectively inserted into sleeves 602, 604 in any desired configuration with respect to which and how many sleeves 602 may receive a neutron moderating member 606. Dynamic neutron reflector assembly 600 substantially surrounds a fuel region 608 from which fast spectrum neutrons 610 emanate. In FIG. 6, lines terminating in double arrows such as lines 612 indicate fast spectrum neutrons. Upon insertion, neutron moderating members 606 displace a volume of flowing neutron reflector material, thus altering the neutron reflectivity characteristics of dynamic neutron reflection assembly 600. Since dynamic neutron reflector assembly 600 contains neutron reflecting and neutron moderating materials, some of the fast spectrum neutrons are reflected back into fuel region 608, and other fast spectrum neutrons 610 strike neutron moderating members 606 and are converted into thermal neutrons. In FIG. 6, example paths of thermal neutrons are indicated by lines terminating in single arrows, such as line 614. Example paths of fast spectrum neutrons are indicated by lines terminating in double arrows. As dynamic reflector assembly 600 converts fast spectrum neutrons into thermal neutrons, the thermal neutrons may be reflected back into fuel region 608 by the flowing neutron reflector material in the channels 602, 604, or reflected by another neutron reflector disposed behind dynamic reflector 600 (not shown). By displacing some of the volume of flowing neutron reflector material, the overall reflectivity characteristics of reflector 600 are changed, thus reducing the breed rate in fuel region 608 due to a reduced reflection of fast spectrum neutrons compared to a configuration without neutron moderating volumetric displacement members 606. The displacement member configuration shown in FIG. 6 also increases the burn rate in fuel region 608 due to an increase in thermal spectrum neutrons compared to a configuration without displacement members. By selectively inserting neutron moderating volumetric displacement members 606 into reflector 600, the breed and burn rates, as well as the neutron spectrum, in fuel region 608 may be dynamically adjusted. The volumetric displacement members 606 may be formed in geometric shapes including without limitation square, round, rectangular, circular, polygonal, etc. In an embodiment, the overall reflectivity characteristics of the reflector 600 are changed by draining one or more of the channels 602, 604 of the flowing neutron reflector material, thus leaving empty space in one or more of the channels 602, 604. Active cooling can be provided to the reflector 600 can provide active cooling by providing thermal communication with the fuel salt and/or with a secondary coolant. In FIG. 6, the neutron reflector material flowing in channels 602 flows upward toward the viewer. In an implementation, neutron reflector material flowing in channels 602 may circulate in channels 602 with input and output ports above fuel region 608 such that no fixtures or ports are needed beneath the reactor. In other implementations, neutron reflector material flowing in channels 602 may flow in only one direction, either in an upward or downward direction, through channels 602 with one port above fuel region 608 and another port below fuel region 608. In yet other implementations, the neutron reflector material may comprise a semi-stagnant or creeping flow through channels 602. In yet other implementations, the neutron reflector material may flow through radial input and output ports. Heat exchanger 614 may be in thermal communication with dynamic reflection assembly 600 for exchanging heat from fuel region 608. In an implementation, the heat exchanger 614 is disposed adjacent on the opposite side of dynamic reflector assembly 600 from fuel region 608. As the neutron reflector material flows through the sleeves of dynamic reflector assembly 600, it may transfer heat emanating from fuel region 608 to the heat exchanger 614 to form a secondary coolant circuit. The secondary coolant circuit may include one or more secondary coolant loops formed from piping. The secondary coolant circuit may include any secondary coolant system arrangement known in the art to be suitable for implementation in a molten fuel salt reactor. The secondary coolant system may circulate a secondary coolant through one or more pipes and/or fluid transfer assemblies of the one or more secondary coolant looks in order to transfer heat generated by the reactor core and received by the heat exchanger 614 to downstream thermally driven electrical generation devices and systems. The secondary coolant system may include multiple parallel secondary coolant loops (e.g., 2-5 parallel loops), each carrying a selected portion of the secondary coolant through the secondary coolant circuit. The secondary coolant may include, but is not limited to, liquid sodium. In an implementation, the heat exchanger 614 is protected by one or more materials effective as a poison or neutron absorber to capture neutrons emanating from the fuel region 608 before the neutrons interact with, and cause radiation damage to, the heat exchanger 614. In an implementation, the heat exchanger 614 includes the one or more materials effective as a poison or neutron absorber. In another implementation, the one or more materials effective as a poison or neutron absorber are included in the dynamic reflector assembly 600. FIG. 7 is a top-down schematic view of a molten nuclear fuel salt fast reactor core with fuel region 702 surrounded by a neutron reflector assembly 700. Neutron reflector assembly 700 contains a neutron reflector material 704 flowing through channels 712. In FIG. 7, neutron reflector material 704 flows upward toward the viewer. In an implementation, neutron reflector material 704 may circulate in channels 712 with input and output ports above fuel region 702 such that no fixtures or ports are needed beneath the reactor. In other implementations, neutron reflector material 704 may flow in only one direction, either in an upward or downward direction, through channels 712 with one port above fuel region 702 and another port below fuel region 702. In yet other implementations, neutron reflector material 704 may comprise a semi-stagnant or creeping flow through channels 712. In yet other implementations, neutron reflector material 704 may flow through radial input and output ports disposed between heat exchangers 706. Flowing dynamic neutron reflector material 704 is in thermal communication with heat exchangers 706. Heat exchangers 706 may contain one or more types of liquid coolant circulating there through. As neutron reflector material 704 exchanges heat with heat exchangers 706, heat exchangers 706 may transport the heat away from neutron reflector assembly 700 as part of a secondary coolant circuit. The secondary coolant circuit may supply heat to electricity generation equipment, such as, for example, a steam-driven turbine. In an implementation, molten fuel salt may flow upward through fuel region 702 and downward through heat exchangers 706, thus exchanging heat as part of a primary coolant circuit. In other words, heat exchangers 706 may exchange heat with both the molten fuel salt and exchange heat with the flowing neutron reflector material 704. The flow rate of neutron reflector material 704 may be adjusted to vary contact time with heat exchangers 706 to vary the temperature of the neutron reflector material 704. As the temperature of the neutron reflector material 704 varies, its density changes accordingly. Changes in the density of neutron reflector material 704 alter its neutron reflective characteristics as denser materials have a higher mass per unit volume and are therefore more likely to reflect neutrons. FIG. 7 shows example fast neutrons 710 emanating from a fuel region 702. Fast neutrons are indicated by lines terminating in double arrows. Example fast neutrons 710 may originate in fuel region 702 and be reflected by a neutron reflector material 704 and travel back into fuel region 702. Example fast neutrons 710 reflected back into fuel region 702 may increase the fissile material in fuel region 702 upon contact with fertile materials. Similarly, FIG. 7 shows example thermal neutrons 714. Example thermal neutrons 714 are indicated by lines terminating in single arrows. Example thermal neutrons 714 may be reflected by neutron reflector material 704 and travel back into fuel region 702. Example thermal neutrons reflected into fuel region 702 may increase the reactivity in fuel region 702 upon contact with fissile material located therein. FIG. 8 is a top-down schematic view of a molten nuclear fuel salt fast reactor core with a fuel region 802 surrounded by a neutron reflector assembly 800 with a neutron reflector material 804 in thermal communication with heat exchangers 806. In FIG. 8, neutron reflector material 804 flows upward toward the viewer. In an implementation, neutron reflector material 804 may circulate in channels 808 with input and output ports above fuel region 802 such that no fixtures or ports are needed beneath the reactor. In other implementations, neutron reflector material 804 may flow in only one direction, either in an upward or downward direction, through channels 808 with one port above fuel region 802 and another port below fuel region 802. In yet other implementations, neutron reflector material 804 may comprise a semi-stagnant or creeping flow through channels 808. In yet other implementations, neutron reflector material 804 may flow through radial input and output ports disposed between heat exchangers 806. Flowing neutron reflector material 804 is in thermal communication with heat exchangers 806. Heat exchangers 806 may contain one or more types of liquid coolant circulating there through. As flowing neutron reflector material 804 exchanges heat with heat exchangers 806, heat exchangers 806 may transport the heat away from the neutron reflector assembly 800 as part of a secondary coolant circuit. The secondary coolant circuit may supply heat to electricity generation equipment, such as, for example, a steam-driven turbine. In an implementation, molten fuel salt may flow upward through fuel region 802 and downward through heat exchangers 806, thus exchanging heat as part of a primary coolant circuit. In other words, heat exchangers 806 may exchange heat with both the molten fuel salt and exchange heat with the flowing neutron reflector material 804. The flow rate of neutron reflector material 804 may be adjusted to vary contact time with heat exchangers 806 to vary the temperature of neutron reflector material 804. The reflector assembly 800 includes neutron moderating volumetric displacement members 812 inserted into fluid channels 808. Upon insertion of moderating members 812, the volume of the reflecting liquid 804 in the channel is reduced. With reduced volume, the remaining neutron reflector material 804 in the channel has an altered neutron reflectivity characteristic, and is therefore less likely to reflect neutrons than before the moderating member 812 was inserted. The presence of moderating member 812 in the area surrounding fuel region 802 makes thermalization of neutrons more likely, such as, for example, thermalized neutron 810. Increased thermalization will tend to increase burnup of fissile material in the fuel region 802. The moderating volumetric displacement members 812 may be inserted into channels 808 singly or in any plurality of members. Moderating volumetric displacement members 812 may take on a cylindrical shape, a square or rectangular prism shape, a triangular prism shape, a polygonal prism shape and the like. In another implementation, moderating volumetric displacement members 812 may include a set of members (not shown). Selection of the geometric shape and number of moderating volumetric displacement members 812 per channel 808 will determine the ratio of moderating material to reflector material in channels 808. Selectively inserting moderating volumetric displacement members 812 permits adjustment of breed rate and reactivity in fuel region 802 and allows maintenance of a desired burnup level. In an implementation, a burnup rate is maintained within a desired upper and lower bound by selectively inserting and removing at least a subset of moderating volumetric displacement members 812. FIG. 9 is a top-down schematic view of a molten nuclear fuel salt fast reactor core with a fuel region 902 surrounded by a neutron reflector assembly 900 with a flowing neutron reflector material 904 through channels 908. In FIG. 9, neutron reflector material 904 flows upward toward the viewer. In an implementation, neutron reflector material 904 may circulate in channels 908 with input and output ports above fuel region 902 such that no fixtures or ports are needed beneath the reactor. In other implementations, liquid neutron reflector 904 may flow in only one direction, either in an upward or downward direction, through channels 908 with one port above the fuel region 902 and another port below the fuel region 902. In yet other implementations, liquid neutron reflector 904 may comprise a semi-stagnant or creeping flow through channels 908. In yet other implementations, liquid neutron reflector 904 may flow through radial input and output ports disposed between heat exchangers 914. Flowing neutron reflector material 904 is in thermal communication with heat exchangers 914. Heat exchangers 914 may contain one or more types of liquid coolant circulating there through. As flowing neutron reflector material 904 exchanges heat with heat exchangers 914, heat exchangers 914 may transport the heat away from neutron reflector assembly 900 as part of a secondary coolant circuit. The secondary coolant circuit may supply heat to electricity generation equipment, such as, for example, a steam-driven turbine. In an implementation, molten fuel salt may flow upward through fuel region 902 and downward through heat exchangers 914, thus exchanging heat as part of a primary coolant circuit. In other words, heat exchangers 914 may exchange heat with both the molten fuel salt and exchange heat with the flowing neutron reflector material 904. The flow rate of neutron reflector material 904 may be adjusted to vary contact time with heat exchangers 914 to vary the temperature of neutron reflector material 904. As the temperature of neutron reflector material 904 varies, its density changes accordingly. Changes in the density of neutron reflector material 904 alter its neutron reflective characteristics as denser liquids have a higher mass per unit volume and are therefore more likely to reflect neutrons. Reflector assembly 900 includes selectively inserted neutron absorbing members 906 and selectively inserted volumetric displacement members 910. Neutron absorbing members 906 and volumetric displacement members 910 may be of any geometric shape compatible with the shape of channels 908. Neutron absorbing members 906 and volumetric displacement members 910 displace a volume of flowing neutron reflector material 904 in the channel 908 into which they are inserted, thus lowering the neutron reflectivity of that channel. Selectively inserting neutron absorbing members 906 and volumetric displacement members 910 adjusts the neutron reflectivity in the nuclear reactor core by altering the composition of the material in the neutron reflection assembly. Several scenarios are possible for fast neutrons travelling into volumetric displacement members 910, such as example fast neutron 910. Fast neutron 912 may pass through the member 910 (not shown in FIG. 9), fast neutron 912 may be reflected by the remaining flowing neutron reflector material 904 in the channel, or fast neutron 912 may be reflected by another surface (not shown). Example fast neutron 912 is less likely to reflect back into fuel region 902 when a volumetric displacement member 910 is inserted than when the channel is full of the flowing neutron reflector material 904. Inserting neutron absorption member 906 is another way of adjusting neutron reflectivity in the nuclear reactor core by altering the composition of the material in the neutron reflection assembly. When neutron absorption member 906 is inserted into a channel 908, example fast neutron 912 may strike and be absorbed by the absorption member 906. Other scenarios are also possible. Example fast neutrons may be reflected by flowing neutron reflector material 904 that was not displaced by absorption member 906, or it may exit the core region where it may be reflected or absorbed by other material (not shown). In another implementation, neutron absorption members 906 may be inserted into a channel 908 while flowing neutron reflector material 904 is removed from the channel. It should be understood that volumetric displacement members 910 and neutron absorption members 906 may be selectively inserted into channels 908 in any desired configuration and in any combination with other members not shown in FIG. 9, such as neutron moderating members. Any number of volumetric displacement members 910 and neutron absorption members 906 may be inserted into a single channel, alone or in combination with other insertable members. Volumetric displacement members 910 and neutron absorption members 906 may be inserted into only some of the channels 908, or only into channels on a portion of dynamic reflector 900. It may be desirable to focus the location of breeding or burning in fuel region 902 by choosing an insertion configuration that concentrates the desired neutron activity in a desired location. For example, an increased breed may be induced in the upper half of fuel region 902 by selectively removing members inserted in the upper half of reflector assembly 900 to allow the neutron reflector material 904 to fill channels 908 on the upper half of the reflector assembly 900. In another example, an increased burn may be induced in the lower half of fuel region 902 by selectively inserting neutron moderating members into the channels 908 on the lower half of reflector assembly 900. In yet another example, reactivity in a portion of fuel region 902 may be reduced by selectively inserting neutron absorbing members 906 into the channels 908 located on the desired side of reflector assembly 900. In the implementation of FIG. 9, flowing neutron reflector material 904 in the channels 908 are in thermal communication with heat exchangers 914. Varying the flow rate of flowing neutron reflector material 904 in channels 908 may alter the flowing reflecting liquid's temperature, and thus its density and thus its neutron reflection characteristics. Altering the density of the flowing neutron reflector material 904 is another way of adjusting the neutron reflectivity in the nuclear reactor core by altering the composition of the material in the neutron reflection assembly. By way of the heat exchangers 914, flowing neutron reflector material 904 in the channels 908 is a secondary coolant for the fuel region 902 because it may operate to exchange heat with the molten fuel salt in the fuel region 902 to the outside of the reactor core via the heat exchangers 914. FIG. 10 is a side schematic view of a molten nuclear fuel salt fast reactor core surrounded by a dynamic neutron reflector assembly 1000 with a neutron reflector material 1002 in thermal communication with a molten nuclear fuel salt 1004 in a tube and shell heat exchanger. Flowing reflecting liquid 1002 flows through inlets 1006 and into outer channels 1008. Outer channels 1008 provide a neutron reflecting layer from which fast neutrons emanating from fuel region 1004 may be reflected back into the fuel region 1004. After leaving outer channels 1012, flowing reflecting liquid 1002 flows through lower channels 1012. Lower channels 1012 provide a neutron reflecting layer from which fast neutrons emanating from fuel salt 1004 may be reflected back into the fuel salt 1004. After leaving lower channels 1012, flowing neutron reflector material 1002 flows upwards through tubes 1014. Tubes 1014 are in thermal communication with molten fuel salt 1004 flowing downward in channels 1016 surrounding tubes 1014 in a shell-and-tube configuration, and therefore function as a secondary coolant for the reactor core. Tubes 1014 may be configured as any number of tubes of any diameter and cross-sectional geometry. Configuration of tubes 1014 may be chosen for a desired surface area contact with flowing molten fuel salt 1004 in the region 1016 for a desired thermal exchange between the flowing neutron reflector material 1002 and the molten fuel salt 1004. After leaving tubes 1014, flowing neutron reflector material 1002 enters upper channel 1020. Upper channel 1020 provides a reflecting layer from which neutrons emanating from fuel region 1004 may be reflected back into fuel region 1004. Heat exchangers (not shown) may be in thermal communication with flowing neutron reflector material 1002. In an implementation, heat exchangers may be disposed outside channel 1008. In another implementation, heat exchangers may be disposed above flowing neutron reflector material inlet 1006 or outlet 1022. By way of the heat exchangers, flowing neutron reflector material 1002 is a secondary coolant for fuel region 1004 because it may operate to exchange heat with the molten fuel salt to the outside of the reactor core. Neutron reflectivity in the nuclear reactor core may be adjusted by altering the composition of the reflecting liquid in channels 1008, 1012, 1020. For example, the volume of flowing neutron reflector material 1002 may be adjusted by pumping an amount of the flowing neutron reflector material 1002 into or out of overflow tank 1010, thus increasing or decreasing the reflectivity, respectively. In another example, the density of flowing neutron reflector material 1002 through channels 1008, 1012, 1020 may be adjusted. A higher density of flowing neutron reflector material 1002 may lead to an increased neutron reflectivity while a lower density of flowing neutron reflector material 1002 may lead to a decreased neutron reflectivity. The density of flowing neutron reflector material 1002 may be adjusted by varying temperature. Temperature of flowing neutron reflector material 1002 may be adjusted by varying flow rate, and thus thermal contact time with the molten fuel salt 1004. Alternatively, or additionally, the direction of flow of the flowing neutron reflector material 1002 may be reversed. As such, the flowing neutron reflector material 1002 may flow in a downward direction through tubes 1014 and up through channels 1008 into overflow tank 1010. The direction of flow of the molten nuclear fuel salt 1004 may also be reversed. As such, the molten nuclear fuel salt 1004 may flow in a downward direction in the center of the fission region and flow in an upward direction around tubes 1014. FIG. 11 is a top-down schematic view of a molten nuclear fuel salt fast reactor core with fuel region 1102 surrounded by a neutron reflector assembly 1100 with a neutron reflector material 1104 flowing through channels 1110, and flowing through tubes 1108 in channels 1112, tubes 1108 being in thermal communication with a molten nuclear fuel salt flowing through fuel region 1102 and through channels 1112 in a tube and shell style heat exchanger. From the viewpoint of FIG. 11, the molten fuel salt flows upward through fuel region 1102 and downward through channels 1112. The flowing reflecting liquid flows downward through channels 1110 and upward through tubes 1108. In this implementation, the flowing reflecting liquid 1104 is also a secondary coolant for the fuel in fuel region 1102. Tubes 1108 may take a variety of configurations, including without limitation any number of tubes in each channel 1112 or tubes of any geometric shape. Selection of the number of tubes 1108 per channel 1112 and the shape of tubes 1108 will determine the surface area in contact with molten fuel salt flowing upward in channel 1112, and alter the amount of heat exchanged between flowing reflecting liquid 1104 and molten fuel salt 1102. Although pairs of tubes 1108 per channel 1112 are shown in FIG. 11, a variety of configurations are possible. For example, tubes 1108 may take a meandering path through channels 1112 to increase surface area thermally exposed to the molten fuel salt. In another implementation, channels 1112 may contain a series of baffles around which the molten fuel salt must flow in an indirect pattern between the inlet and outlet ports. The indirect flow pattern increases the thermal contact between the molten fuel salt and the tubes, and increases the angle between the tubes and the molten fuel salt flow to increase thermal communication. In an embodiment, example fast neutrons 1114 emanating from fuel region 1102 may be reflected by flowing reflecting liquid 1104 contained in tubes 1008 or be reflected by flowing reflecting liquid 1104 contained in channels 1110, and back into fuel region 1102. Fast neutrons such as example fast neutron 1116 emanating from molten fuel salt flowing in channels 1112 may also be reflected by flowing reflector material 1104 in tubes 1108 or in channels 1110, and back into fuel region 1102. FIG. 12 depicts a flow diagram of example operations 1200 of dynamic spectrum shifting in a molten nuclear fuel salt fast reactor. A sustaining operation 1202 sustains a nuclear fission reaction in a nuclear reactor core surrounded by a dynamic neutron reflector assembly. The neutron reflector assembly may have at least one neutron reflector material. A neutron reflection assembly may surround a nuclear reactor core by being disposed radially around, above, and/or below the reactor core. The neutron reflection assembly may be formed in one contiguous piece, formed into discrete pieces distributed around the reactor core, disposed around the core in discrete pieces with gaps in between, and/or segmented into regular or irregular sections. The reflection assembly may contain one or more channels for conducting a flowing reflector material. The reflection assembly may contain one or more levels of channels, such that a flowing reflector material flows in one direction in one level, and flows in another direction in one or more other levels. For example, the reflection assembly may contain an outer channel with flowing reflector material flowing downward, and another inner channel with flowing reflector material flowing upward to avoid any inlet or outlet plumbing underneath the reactor core. The reflection assembly may further be in thermal communication with one or more heat exchangers, and therefore function as a secondary coolant for the reactor core. In one implementation, heat exchangers are thermally coupled to channels for conducting the flowing reflector material. Another implementation may utilize a tube-and-shell heat exchanger wherein a first channel conducts a flowing reflector material in a first direction, and one or more additional channels conduct the flowing reflector material in a second direction through one or more tubes surrounded by flowing molten fuel salt. An adjusting operation 1204 adjusts fast neutron flux and thermal neutron flux within the nuclear reactor core during the sustained nuclear fission reaction by altering reflectivity characteristics of reflector material in the neutron reflector assembly. Altering reflectivity characteristics of reflector material in the neutron reflector assembly may include: any one or more of modifying the volume of reflector material in the reflector assembly, modifying the density of reflector material in the reflector assembly, modifying the composition of reflector material in the reflector assembly, inserting and/or removing neutron moderating members into the reflector assembly, inserting and/or removing neutron absorbing members into the reflector assembly, and/or inserting and/or removing volumetric displacement members into the reflector assembly. FIG. 13 depicts a flow diagram of other example operations 1300 of dynamic spectrum shifting in a molten nuclear fuel salt fast reactor. A sustaining operation 1302 sustains a nuclear fission reaction in a nuclear reactor core surrounded by a neutron reflector assembly. The neutron reflector assembly may have at least one neutron reflector material. A neutron reflection assembly may surround a nuclear reactor core by being disposed radially around, above, and/or below the reactor core. The neutron reflection assembly may be formed in one contiguous piece, formed into discrete pieces distributed around the reactor core, disposed around the core in discrete pieces with gaps in between, and/or segmented into regular or irregular sections. The reflection assembly may contain one or more channels for conducting a flowing reflector material. The reflection assembly may contain one or more levels of channels, such that a flowing reflector material flows in one direction in one level, and flows in another direction in one or more other levels. For example, the reflection assembly may contain an outer channel with flowing reflector material flowing downward, and another inner channel with flowing reflector material flowing upward to avoid any inlet or outlet plumbing underneath the reactor core. The reflection assembly may further be in thermal communication with one or more heat exchangers, and therefore function as a secondary coolant for the reactor core. In one implementation, heat exchangers are thermally coupled to channels for conducting the flowing reflector material. Another implementation may utilize a tube-and-shell heat exchanger wherein a first channel conducts a flowing reflector material in a first direction, and one or more additional channels conduct the flowing reflector material in a second direction through one or more tubes surrounded by flowing molten fuel salt. An adjusting operation 1304 adjusts fast neutron flux and thermal neutron flux within the reactor core during the sustained nuclear fission reaction by modifying the volume of reflector material in the neutron reflector assembly. In an implementation, volume of a flowing reflector material may be altered by a pump and valve fluidically coupled to a spillover reservoir. A volume of flowing reflector material may be pumped through the valve and into the spillover reservoir to reduce volume of reflector material in the reflection assembly, and thus reduce the flux of fast and/or thermal neutrons scattered into the reactor core. Conversely, a volume of flowing material may be pumped though the valve out of the spillover reservoir to increase volume in the reflector assembly, and thus increase reflectivity of neutrons into the reactor core. In another implementation, altering the composition of material in the neutron reflector assembly may include selectively inserting or removing a volumetric displacement member into one or more channels conducting a flowing reflector material. In implementations, a volumetric displacement member may be a neutron moderating member, a neutron absorbing member, or a volumetric displacement member that does not influence neutron flux (e.g., a hollow member or a member formed of non-neutron influencing materials). Insertion of a volumetric displacement member into a channel conducting flowing reflector material surrounding a reactor core reduces the volume of the reflector material in a channel, and thus alters the reflectivity characteristics of the reflector assembly by reducing the scattering of neutrons because fewer neutrons are likely to be scattered due to a reduced volume of reflector material. Removing a volumetric displacement member from a channel conducting a flowing reflector material surrounding a nuclear reactor core may increase the volume of the flowing reflector material, and thus alters the reflectivity characteristics of the reflector assembly by increasing the scattering of neutrons because flowing reflector material may return to the reflector assembly into the space vacated by the withdrawn volumetric displacement member, thus increasing the likelihood that neutrons emanating from a reactor core will be scattered due to increased volume of reflector material. FIG. 14 depicts a flow diagram of other example operations 1400 of dynamic spectrum shifting in a molten nuclear fuel salt fast reactor. A sustaining operation 1402 sustains a nuclear fission reaction in a nuclear reactor core surrounded by a neutron reflector assembly. The neutron reflector assembly may have at least one neutron reflector material. A neutron reflection assembly may surround a nuclear reactor core by being disposed radially around, above, and/or below the reactor core. The neutron reflection assembly may be formed in one contiguous piece, formed into discrete pieces distributed around the reactor core, disposed around the core in discrete pieces with gaps in between, and/or segmented into regular or irregular sections. The reflection assembly may contain one or more channels for conducting a flowing reflector material. The reflection assembly may contain one or more levels of channels, such that a flowing reflector material flows in one direction in one level, and flows in another direction in one or more other levels. For example, the reflection assembly may contain an outer channel with flowing reflector material flowing downward, and another inner channel with flowing reflector material flowing upward to avoid any inlet or outlet plumbing underneath the reactor core. The reflection assembly may further be in thermal communication with one or more heat exchangers, and therefore function as a secondary coolant for the reactor core. In one implementation, heat exchangers are thermally coupled to channels for conducting the flowing reflector material. Another implementation may utilize a tube-and-shell heat exchanger wherein a first channel conducts a flowing reflector material in a first direction, and one or more additional channels conduct the flowing reflector material in a second direction through one or more tubes surrounded by flowing molten fuel salt. An adjusting operation 1404 adjusts fast neutron flux and thermal neutron flux within the reactor core during the sustained nuclear fission reaction by modifying the density of reflector material in the neutron reflector assembly. Density of reflector material in the neutron reflector assembly may be modified by altering the temperature of a flowing neutron reflector material in the reflector assembly. At higher temperatures, a flowing neutron reflector material tends to have lower density, and, at lower temperatures, a flowing neutron reflector material tends to have higher density. Changes in density will alter the alter the reflectivity characteristics of the reflector assembly because fast and thermal neutrons emanating from the reactor core will be more or less likely to be scattered by the reflector material depending on the likelihood of a collision with the nuclei of the reflector material in the reflector assembly. One way of altering the temperature of a flowing neutron reflector material is to alter its flow rate, and thus the thermal contact time the flowing reflector material has with a molten fuel salt. A higher flow rate may reduce contact time with a hot fuel salt, thus lowering the flowing reflector material's temperature and increasing the flowing reflector material's density. A lower flow rate may leave the flowing reflector material in thermal contact with the hot fuel salt for a relatively longer period of time, thus increasing its temperature and lowering the flowing reflector material's density. In another embodiment, a tube and shell heat exchanger may be employed to exchange heat between the flowing reflector material and the molten fuel salt. The tube and shell heat exchanger may be configured with baffles to route the molten fuel salt in a meandering path around tubes carrying the flowing reflector material. Movable baffles may increase or decrease the thermal contact time between the flowing reflector material and the molten fuel salt. As described above, a change in thermal contact time between the flowing reflector material and the molten fuel salt may tend to alter the temperature, and thus density, of the flowing reflector material. FIG. 15 depicts a flow diagram of other example operations 1500 of dynamic spectrum shifting in a molten nuclear fuel salt fast reactor. A sustaining operation 1502 sustains a nuclear fission reaction in a nuclear reactor core surrounded by a dynamic neutron reflector assembly. The neutron reflector assembly may have at least one neutron reflector material. A neutron reflection assembly may surround a nuclear reactor core by being disposed radially around, above, and/or below the reactor core. The neutron reflection assembly may be formed in one contiguous piece, formed into discrete pieces distributed around the reactor core, disposed around the core in discrete pieces with gaps in between, and/or segmented into regular or irregular sections. The reflection assembly may contain one or more channels for conducting a flowing reflector material. The reflection assembly may contain one or more levels of channels, such that a flowing reflector material flows in one direction in one level, and flows in another direction in one or more other levels. For example, the reflection assembly may contain an outer channel with flowing reflector material flowing downward, and another inner channel with flowing reflector material flowing upward to avoid any inlet or outlet plumbing underneath the reactor core. The reflection assembly may further be in thermal communication with one or more heat exchangers, and therefore function as a secondary coolant for the reactor core. In one implementation, heat exchangers are thermally coupled to channels for conducting the flowing reflector material. Another implementation may utilize a tube-and-shell heat exchanger wherein a first channel conducts a flowing reflector material in a first direction, and one or more additional channels conduct the flowing reflector material in a second direction through one or more tubes surrounded by flowing molten fuel salt. An adjusting operation 1504 adjusts fast neutron flux and thermal neutron flux within the reactor core during the sustained nuclear fission reaction by inserting a neutron moderating member into the neutron reflector assembly. Insertion of a neutron moderating member may introduce nuclei into the reflector assembly that may tend to cause elastic collisions with fast neutrons. The presence of these nuclei may scatter thermal neutrons back into the nuclear reactor core, thus increasing burnup. Adjusting operation 1504 may also have an effect on the neutron reflectivity characteristics of the neutron reflection assembly because the neutron moderating member will displace a volume of flowing neutron reflector material from the neutron reflector assembly. The decrease in volume of flowing neutron reflector material will tend to decrease the amount of elastic collisions with neutrons emanating from the nuclear reactor core, thus reducing the likelihood of scattering fast neutrons emanating from the nuclear reactor core back into the reactor core to breed fertile material into fissile material. FIG. 16 depicts a flow diagram of other example operations 1600 of dynamic spectrum shifting in a molten nuclear fuel salt fast reactor. A sustaining operation 1602 sustains a nuclear fission reaction in a nuclear reactor core surrounded by a dynamic neutron reflector assembly. The neutron reflector assembly may have at least one neutron reflector material. A neutron reflection assembly may surround a nuclear reactor core by being disposed radially around, above, and/or below the reactor core. The neutron reflection assembly may be formed in one contiguous piece, formed into discrete pieces distributed around the reactor core, disposed around the core in discrete pieces with gaps in between, and/or segmented into regular or irregular sections. The reflection assembly may contain one or more channels for conducting a flowing reflector material. The reflection assembly may contain one or more levels of channels, such that a flowing reflector material flows in one direction in one level, and flows in another direction in one or more other levels. For example, the reflection assembly may contain an outer channel with flowing reflector material flowing downward, and another inner channel with flowing reflector material flowing upward to avoid any inlet or outlet plumbing underneath the reactor core. The reflection assembly may further be in thermal communication with one or more heat exchangers, and therefore function as a secondary coolant for the reactor core. In one implementation, heat exchangers are thermally coupled to channels for conducting the flowing reflector material. Another implementation may utilize a tube-and-shell heat exchanger wherein a first channel conducts a flowing reflector material in a first direction, and one or more additional channels conduct the flowing reflector material in a second direction through one or more tubes surrounded by flowing molten fuel salt. An adjusting operation 1604 adjusts fast neutron flux and thermal neutron flux within the reactor core during the sustained nuclear fission reaction by removing a neutron moderating member out of the neutron reflector assembly. Removal of a neutron moderating member will reduce available nuclei in the reflector assembly that may tend to cause elastic collisions with fast neutrons. The reduced presence of these nuclei will scatter fewer thermal neutrons back into the nuclear reactor core, thus decreasing burnup. Adjusting operation 1504 may also have an effect on the neutron reflectivity characteristics of the neutron reflection assembly because the removed neutron moderating member may have displaced a volume of flowing neutron reflector material when it had been inserted in the neutron reflector assembly. An increase in volume of flowing neutron reflector material may tend to increase the amount of elastic collisions with neutrons emanating from the nuclear reactor core, thus increasing the likelihood of scattering fast neutrons emanating from the nuclear reactor core back into the reactor core to breed fertile material into fissile material. FIG. 17 depicts a top-down schematic view of an example neutron reflector assembly 1700. Neutron reflector assembly 1700 includes two sub-assemblies, a primary static neutron reflector sub-assembly 1712 and a secondary dynamic neutron reflector sub-assembly 1716. In FIG. 17, example paths of fast spectrum neutrons 1706, 1714 are indicated by lines terminating in double arrows, such as lines 1708 indicate example fast spectrum neutrons. In an implementation, flowing neutron reflector assembly 1700 substantially surrounds a nuclear fuel region 1704 from which fast spectrum neutrons 1706, 1714 emanate. Primary static neutron reflector sub-assembly 1712 may contain a neutron reflector material. The neutron reflector material contained in primary static neutron reflector sub-assembly 1712 may be a solid, liquid, or fluid neutron reflector material, or a combination thereof. The primary static neutron reflector sub-assembly 1712 may substantially surround a fuel region 1704. In another implementation, primary static neutron reflector sub-assembly 1712 may partially surround the fuel region 1704 in a continuous, segmented, and/or modular manner. The example fast spectrum neutrons 1714 emanating from nuclear fuel region 1704 are inelastically scattered (or reflected) from the primary static neutron sub-assembly 1716 and back into the nuclear fuel region 1704, thus increasing a breed rate of fertile fuel in the fuel region 1704. Other example fast spectrum neutrons, such as example neutrons 1706 may pass through primary static neutron reflector sub-assembly 1712, and be inelastically scattered (or reflected) from secondary dynamic neutron reflector sub-assembly 1716, as explained in more detail below. The primary static neutron reflector sub-assembly 1712 may be disposed adjacent to, and/or in thermal contact with, the nuclear fuel region 1704. Due to the positioning of primary static neutron sub-assembly 1712 with respect to the nuclear fuel region 1704, the primary static neutron reflector sub-assembly 1712 may experience high levels of exposure to forces that may cause damage or wear. For example, the primary static neutron reflector sub-assembly may be exposed to high levels of heat and various types of radiation emanating from the nuclear fuel region 1704, including without limitation, alpha particles, beta particles, and/or gamma rays. Prolonged exposure to heat and/or radiation may cause the primary static neutron reflector sub-assembly 1712 to suffer excessive structural degrading over a period of time. The primary static neutron reflector sub-assembly 1712 may therefore be removable from flowing neutron reflector assembly 1700. In other words, the primary static neutron reflector sub-assembly may, or modular parts thereof, may be slidably fitted to a housing (not shown) to permit selective replacement of the sub-assembly, which may be carried out according to a periodic maintenance schedule or based on periodic inspection of the primary static neutron reflector sub-assembly 1712. FIG. 17 also illustrates a secondary dynamic neutron reflector sub-assembly 1716. Secondary dynamic neutron reflector sub-assembly 1716 may be used to incrementally shift neutron spectrum in nuclear fuel region 1704 by selectively filling each of the channels 1702 with a volume of neutron reflector material. Secondary dynamic neutron reflector assembly 1716 may include a plurality of refractory-clad sleeves 1702 to conduct a flowing neutron reflector material there through. In FIG. 17, the neutron reflector material flows upward through a refractory clad channel 1702 toward the viewer. In an implementation, neutron reflector material may circulate in channels 1702 (e.g., cells, sleeves, conduits, etc.) with input and output ports above the nuclear fuel region 1704 such that no fixtures or ports are needed beneath the reactor. In other implementations, the neutron reflector material may flow in only one direction, either in an upward or downward direction, through the channels 1702 with one port above the nuclear fuel region 1704 and another port below fuel region 1704. In yet other implementations, the neutron reflector material may comprise a semi-stagnant or creeping flow through the channels 1702. In yet other implementations, the neutron reflector material may flow through radial input and output ports. The secondary dynamic neutron reflector sub-assembly 1716 is in thermal communication with heat exchanger 1710 disposed on the opposite side from fuel region 1704. It is to be appreciated that the dynamic neutron reflector assembly and/or the heat exchanger could be inside, or disposed among the static reflector sub-assembly. The heat exchanger 1710 may contain one or more types of liquid coolant circulating there through. As secondary dynamic neutron reflector sub-assembly 1716 exchanges heat with the heat exchanger 1710, the heat exchanger 1710 may transport the heat away from the secondary dynamic neutron reflector sub-assembly 1716 as part of a secondary coolant circuit. The secondary coolant circuit may supply heat to electricity generation equipment, such as, for example, a steam-driven turbine. In an implementation, molten fuel salt may flow upward through the nuclear fuel region 1704 and downward through the heat exchanger 1710, thus exchanging heat as part of a primary coolant circuit. In other words, the heat exchangers may exchange heat with both the molten fuel salt and exchange heat with the flowing neutron reflector in the channels 1702. The flow rate of neutron reflector material may be adjusted to vary contact time with the heat exchangers to vary the temperature of reflector material flowing in the channels 1702. As the temperature of reflector material varies, its density changes accordingly. Changes in the density of the reflector material alter its neutron reflective characteristics as denser materials have a higher mass per unit volume and are therefore more likely to reflect neutrons. FIG. 18 is a top-down schematic view of a molten nuclear fuel salt fast reactor core with a fuel region 1802 surrounded by a neutron reflector assembly 1800. The neutron reflector assembly includes an inner annular channel 1808 and outer annular channel 1810 surrounding fuel region 1802. The inner and outer annular channels 1808, 1810 may contain neutron reflector materials 1804 and 1806, respectively. The neutron reflector materials 1804, 1806 may be the same or differ from one another in terms of their respective neutron-reflecting properties or other properties that may affect performance of the neutron reflector assembly (viscosity, density, specific heat value, etc.). Neutron reflector materials 1804, 1806 may tend to reflect example fast neutrons 1812 back into fuel region 1802. Neutron reflector materials 1804, 1806 may be selectively added, removed, and/or replaced in channels 1808, 1810 to dynamically alter the neutron reflecting characteristics of the neutron reflector assembly 1800 over time. In one implementation, one or both of the neutron reflector materials 1804, 1806 may be completely removed from their respective channels 1808, 1810 to alter the neutron reflecting characteristics of the neutron reflector assembly 1800. In another implementation, the neutron reflector materials 1804, 1806 may be the same material. In yet another implementation, the neutron reflector materials 1804, 1806 may be selectively added, removed, and/or replaced to provide lower neutron reflection near the beginning of the life of the reactor when there is greater breeding of fertile fuel, and selectively added, removed, and/or replaced to provide greater neutron reflection as the reactor ages and burnup begins to dominates in the fuel region 1802. In another implementation, neutron reflector materials 1804, 1806 may mix inside one or both of channels 1808, 1810. In yet another implementation, one or both of neutron reflector materials 1804, 1806 may be added over time to channels 1808, 1810 to alter the ratio between the two materials and thus the neutron reflectivity of the assembly. If more than two neutron reflector materials 1804, 1806 are mixed inside channels 1808, 1810, a separator component (not shown) may operate to separate the materials if desired and may operate in any suitable manner to separate the two or more neutron reflector materials including one or more suitable chemical, mechanical, magnetic, electrical, time-bases processes based on the chemical and physical properties of the two or more neutron reflector materials. In another embodiment, mixed neutron reflector materials 1804, 1806 may be separated via a flush operation. Alternatively, the neutron reflector materials 1804, 1806 may be held in separate reservoirs (not shown) to selectively source the flows into one or both of channels 1808, 1810. In an implementation, neutron reflector materials 1804, 1806 may circulate in channels 1808, 1810 with input and output ports above fuel region 1802 such that no fixtures or ports are needed beneath the reactor. In other implementations, neutron reflector materials 1804, 1806 may flow in only one direction, either in an upward or downward direction, through channels 1808, 1810 with one port above fuel region 1802 and another port below fuel region 1802. In yet other implementations, neutron reflector materials 1804, 1806 may comprise a semi-stagnant or creeping flow through channels 1808, 1810. In yet other implementations, neutron reflector materials 1804, 1806 may flow through radial input and output ports. In another implementation, the channels 1808, 1810 may be selectively filled with materials that are not neutron reflectors. In one example, the channels 1808, 1810 may be filled with neutron moderating materials, neutron absorbing materials, or neutronically translucent materials. In another implementation, one or both of the channels 1808, 1810 may include selectively insertable volumetric displacement members 1814. Volumetric displacement members 1814 may contain neutron moderating materials, neutron absorbing materials, or neutronically translucent materials. Upon insertion of volumetric displacement members 1814, the volume of the reflecting liquid 1804, 1806 in the channel into which the volumetric displacement member has been inserted is reduced. With reduced volume, the remaining neutron reflector material 1804, 1806 in the channel has an altered neutron reflectivity characteristic, and is therefore less likely to reflect neutrons than before the volumetric displacement member 1814 was inserted. FIG. 19 is a top-down schematic view of a molten nuclear fuel salt fast reactor core with a fuel region 1902 surrounded by a neutron reflector assembly 1900. The neutron reflector assembly includes an inner annular channel 1908 and outer annular channel 1910 surrounding fuel region 1902. The inner and outer annular channels 1908, 1910 may contain a neutron reflector material 1904. In an implementation, neutron reflector material 1904 may circulate in channels 1908, 1910 with input and output ports above fuel region 1902 such that no fixtures or ports are needed beneath the reactor. In other implementations, neutron reflector material 1904 may flow in only one direction, either in an upward or downward direction, through channels 1908, 1910 with one port above fuel region 1902 and another port below fuel region 1902. In yet other implementations, neutron reflector material 1904 may comprise a semi-stagnant or creeping flow through channels 1908, 1910. In yet other implementations, neutron reflector material 1904 may flow through radial input and output ports. In one implementation, neutron reflector material 1904 may flow through channels 1908, 1910 at time periods near the beginning of the life of the reactor with fuel region 1902. As the reactor breeds fertile fuel over time, the effectiveness of the neutron reflector assembly 1900 may decrease because the inventory of bred nuclear fuel may exceed the amount needed to fuel the reactor. It may be desirable to therefore replace a portion of the neutron reflector material in part of the neutron reflector assembly as shown in FIG. 20 to alter the shape of the neutron reflector assembly 1900 over time. FIG. 20 is a top-down schematic view of a molten nuclear fuel salt fast reactor core with a fuel region 2002 surrounded by a neutron reflector assembly 2000. In FIG. 20 the neutron reflector material contents of inner annular channels 2008 are selectively replaced with additional fuel salt from fuel region 2002. As a result, the reactor will experience less neutron “leak.” Example fast neutrons 2012 may continue to experience reflection against neutron reflection material 2006 in channel 2010. It is therefore possible to start a fission reaction in the reactor core with a smaller volume of fuel salt near the beginning of the life of the reactor because more fissile fuel materials may be bred as the reactor operates. The additional bred fuel may replace a volume of neutron reflector material in the channels 2008. This may lower the upfront cost of operating the reactor and enhance the breeding of the reactor later in life when breeding is more challenging due at least in part to built-up fission products. Neutron reflector materials 2006 may tend to reflect example fast neutrons 2012 back into the fuel salt, whether the example fast neutrons 2012 emanate from fuel region 2002 or from inner annular channels 2008. FIG. 21 is a top-down schematic view of a molten nuclear fuel salt fast reactor core with a fuel region 2102 surrounded by a neutron reflector assembly 2100. The neutron reflector assembly includes a plurality of annular channels 2104 surrounding the fuel region 2102. The annular channels 2104 may contain a plurality of tubes 2108 containing a flowing neutron reflector material 2106 in neutronic communication with the fuel region 2102. In an implementation, the plurality of tubes 2108 are cylindrical tubes. The flowing neutron reflector material 2106 may be circulated in the tubes 2108 with input and output ports above fuel region 2102 such that no fixtures or ports are needed beneath the reactor. In other implementations, neutron reflector material 2106 may flow in only one direction, either in an upward or downward direction, through the tubes 2108 with one port above fuel region 2102 and another port below fuel region 2102. In yet other implementations, neutron reflector material 2106 may comprise a semi-stagnant or creeping flow through tubes 2108. The tubes 2108 are arranged such that the radius of all tubes 2108 is not equal. As such, a plurality of tubes 2108 with varying radius values may be disposed in a channel 2104. In an implementation, tubes 2108 of varying radius may flow neutron reflector material in a volume that occupies a cross-sectional area of 80% of the cross-sectional area of the channels 2104. Numerals have not been assigned to every tube to improve readability due to the large number of tubes 2108 depicted in FIG. 21. This disclosure should be understood as indicating that each tube shown in channels 2104 is a tube 2108 containing neutron reflector material 2106, even those that are not so numbered therein. As discussed above, in some embodiments reflectors or portions of reflectors may be completely solid at operating temperatures, e.g., between 300-350° C. and 800° C., or could be a liquid reflector material encased in an enclosed container in which the container walls are solid at operating temperature. Examples of solid reflector materials include uranium, uranium-tungsten, carbides of uranium or uranium-tungsten, and magnesium oxide. Examples of reflector materials that could be used as a liquid coolant include lead, lead alloys, PbBi eutectic, PbO, iron-uranium alloys including iron-uranium eutectic, graphite, tungsten carbide, densalloy, titanium carbide, depleted uranium alloys, tantalum tungsten, and tungsten alloys. In yet another embodiment fuel salt may be used as reflector material. In an embodiment, liquid coolant includes materials that are liquid at the reactor operating temperature and that have a density greater than 10 grams/cm3. In an alternative embodiment, liquid coolant includes materials that are liquid at the reactor operating temperature and that exhibit an elastic cross section of 0.1 barns or greater for 0.001 MeV neutrons. As discussed above, examples of liquid nuclear fuels include salts containing one or more of PuCl3, UCl4, UCl3F, UCl3, UCl2F2, UClF3, bromide fuel salts such as UBr3 or UBr4, and thorium chloride (e.g., ThCl4) fuel salts. Furthermore, a fuel salt may include one or more non-fissile salts such as, but not limited to, NaCl, MgCl2, CaCl2, BaCl2, KCl, SrCl2, VCl3, CrCl3, TiCl4, ZrCl4, ThCl4, AcCl3, NpCl4, AmCl3, LaCl3, CeCl3, PrCl3 and/or NdCl3. Note that the minimum and maximum operational temperatures of fuel within a reactor may vary depending on the fuel salt used in order to maintain the salt within the liquid phase throughout the reactor. Minimum temperatures may be as low as 300-350° C. and maximum temperatures may be as high as 1400° C. or higher. Similarly, except were explicitly discussed otherwise, heat exchangers will be generally presented in this disclosure in terms of simple, single pass, shell-and-tube heat exchangers having a set of tubes and with tube sheets at either end. However, it will be understood that, in general, any design of heat exchanger may be used, although some designs may be more suitable than others. For example, in addition to shell and tube heat exchangers, plate, plate and shell, printed circuit, and plate fin heat exchangers may be suitable. FIG. 22 illustrates a cross-section view of an embodiment of a reactor 2200 utilizing a circulating reflector material. The illustration shows the half of the reactor 2200 from the center to the left edge of the containment vessel 2218. The reactor 2200 includes a reactor core 2204 defined by an upper reflector 2208A, a lower reflector 2208B and an inner reflector 2208C. In the embodiment shown, the lower reflector 2208B also extends laterally and up the sides of the containment vessel 2218 for added protection to the vessel head 2238. The primary heat exchanger 2210 configured to have shell-side coolant flow (illustrated by dotted lines 2214), the coolant entering through a coolant inlet channel 2230 and heated coolant exiting from coolant outlet channel 2236. In the embodiment shown, fuel flows (illustrated by dashed lines 2206) from the reactor core 2204, via an upper channel through the inner reflector 2208C, and into the heat exchanger 2210 through the inlet tube sheet 2232. After passing through the tube set, the now-cooled fuel exits the lower tube sheet 2231 and flows back into the reactor core 2204 via a lower channel through the inner reflector 2208C. Flow of the fuel is driven by a pump assembly 2212 that includes an impeller in the fuel circuit (in this embodiment illustrated below the lower tube sheet 2231) connected by a shaft to a motor (in this embodiment located above the upper reflector 2208A). In FIG. 22, the reflectors 2208A, 2208B, 2208C are in fluid communication allowing liquid reflector material to be circulated around the reactor core 2204. Flow of the reflector material is illustrated in FIG. 22 by the large, gray arrows 2234. In the embodiment shown, reflector material flows through an inlet in the vessel head 2238 into reactor 2200 along the interior surface of the containment vessel 2218 and then along the bottom of the containment vessel 2218 before rising and making a U-turn to flow adjacent to the bottom of the reactor core 2204. The reflector material then flows up through the inner reflector 2208C then into the upper reflector 2208A from which it can be removed via an outlet in the vessel head 2238 or recirculated to the interior surface of the containment vessel 2218. The circulating reflector material in FIG. 22 may be used to assist in the cooling of the reactor core 2204. In this configuration, the heated reflector material may be removed from the containment vessel 2218 and passed through a heat exchanger 2299 external to the reactor 2200. In an embodiment, the same primary coolant loop that removes heat directly from the fuel via heat exchanger 2210 may also be used to remove heat from the reflector material. In an alternative embodiment, a separate and independent cooling system may be used to remove the heat from the reflector material which may use the same type of coolant as the primary coolant or a different type of coolant. In yet another embodiment, the reflector material cooling may be incorporated into an auxiliary cooling system that provides emergency cooling to the reflector material in the event of a loss of flow in the primary cooling loop. In the embodiment shown, when the reflector material is part of a cooling loop, a benefit of the configuration illustrated in FIG. 22 is that the containment vessel is both actively cooled and protected from excessive neutron flux. Because cooled reflective material is first flowed along the interior surfaces of the containment vessel 2218 prior to flowing to locations near the reactor core 2204, the initial temperature of the cooled reflective material can be used control the temperature of the containment vessel 2218. In yet another embodiment, a cooling jacket (not shown) can be provided on the exterior surface of the containment vessel 2218, which serves to remove heat from the circulating reflective material on the interior surface of the containment vessel 2218. This may be done in addition to or instead of an exterior reflective material cooling circuit. As described above, the overall reflectivity of the reflector configuration of FIG. 22 may be controlled by controlling the flow rate of reflective material through the reflectors as well as by inserting or removing rods or other components containing moderating materials or materials of different reflectivity from that of the circulating reflective material. FIG. 23 illustrates an embodiment of a reactor with a shell-side fuel/tube-side primary coolant heat exchanger configuration using the same cross-section view of half of the reactor as in FIG. 22. The reactor core 2304 is surrounded by an upper reflector 2308A, a lower reflector 2308B, and an inner reflector 2308C that separates the reactor core from the primary heat exchanger 2310. The channels are provided through the reflectors 2308A, 2308B, 2308C allowing the circulation of fuel salt (illustrated by a dashed line 2306) from the reactor core 2304 through the inner reflector 2308C, into the shell of the primary heat exchanger 2310. The fuel flows through the shell around the tube set, thus transferring heat to the primary coolant. Cooled fuel then exits the shell and passes through the inner reflector 2308C back into the bottom of the reactor core 2304. Baffles 2312 are provided in the shell to force the fuel salt to follow a circuitous path around the tubes of the heat exchanger for more efficient heat transfer. Coolant flows through the tube-side of the heat exchanger 2310, but before entering the bottom of the heat exchanger first flows through an inlet in the vessel head 2338, down the length of a coolant inlet channel 2330 adjacent to a portion of the lower reflector 2308B. The primary coolant enters the tubes of the heat exchanger 2310 by flowing through the lower tube sheet 2331, which is illustrated as being level with the bottom of the reactor core. The lower tube sheet 2331 may be at or below the level of the lower reflector 2308B depending on the embodiment. The coolant exits the tubes of the heat exchanger at the upper tube sheet 2332, which is located in FIG. 23 some distance above the reactor core 2304 and containment vessel 2318. The flow of the coolant is also illustrated by a dashed line 2314. FIG. 23 illustrates a region 2334 within the shell of the heat exchanger that is above the level of salt in the reactor core 2304. This region may either be solid, except for the penetrating tubes, or may be a headspace filled with inert gas. One or more pumps (not shown) may be provided to assist in the fuel salt circulation, the primary coolant circulation or both. For example, an impeller may be provided in one or both of the heated fuel salt inlet channel at the top of the reactor core 2304 or (as discussed in greater detail below) the cooled fuel outlet channels at the bottom of the reactor core 2304. Likewise, an impeller may be provided in the coolant inlet channel 2330 to assist in control of the primary coolant flow. In FIG. 23, the reflectors 2308A, 2308B, 2308C are in fluid communication allowing liquid reflector material to be circulated around the reactor core 2304. Flow of the reflector material is illustrated in FIG. 23 by the large, gray arrows 2334. In the embodiment shown, reflector material flows into reactor 2300 through an inlet in the vessel head 2338 and then along the interior surface of the side of the containment vessel 2318 in a reflector channel. The reflector channel then follows the bottom of the containment vessel 2318 before making a U-turn and rising to flow adjacent to the bottom of the reactor core 2304. The reflector material then flows up through the inner reflector 2308C and into the upper reflector 2308A from which it can be removed at a central location via an outlet in the vessel head 2338, as shown, or recirculated to the interior surface of the containment vessel 2318. As discussed with reference to FIG. 22, the circulating reflector material in FIG. 23 may be used to assist in the cooling of the reactor core 2304. In this configuration, the heated reflector material may be removed from the containment vessel 2318 and passed through a heat exchanger (not shown) external to the reactor 2300. When the reflector material is part of a cooling loop, a benefit of the configuration illustrated in FIG. 23 is that the containment vessel is both actively cooled and protected from excessive neutron flux. Because cooled reflective material is first flowed along the interior surfaces of the containment vessel 2318 prior to flowing to locations near the reactor core 2304, the initial temperature of the cooled reflective material can be used control the temperature of the containment vessel 2318. As described above, the overall reflectivity of the reflector configuration of FIG. 23 may be controlled by controlling the flow rate of reflective material through the reflectors as well as by inserting or removing rods or other components containing moderating materials or materials of different reflectivity from that of the circulating reflective material. As discussed above, yet another approach to cooling the reactor is to utilize a liquid reflector as the primary coolant. In this design, the primary coolant performs both the function of the reflectors and the primary cooling functions. In an embodiment, a reflector material will be liquid at the minimum operational fuel salt temperature (for example, between 300° C. and 800° C.) and have a density greater than 10 grams/cm3. In an alternative embodiment, a reflector material may be a material having a low neutron absorption cross section and a high scattering cross section and that may undergo (n,2n) reactions. FIG. 24 illustrates such an embodiment of a reflector cooled reactor. In the embodiment, half of the reactor 2400 is illustrated in cross-section as in FIGS. 22 and 23. The reactor core 2404 is surrounded by an upper reflector 2408A, a lower reflector 2408B. Molten reflector material, such as lead, flowing through the coolant inlet channel as illustrated by gray arrow 2414 acts as the inner reflector 2408C as well as the primary coolant. Any type of system may be used to circulate the reflector material. In the embodiment in FIG. 24, for example, a pump 2413 as described with reference to FIG. 22 is provided in the cooled material inlet channel. Such a pump 2413 may be located so that the impeller is at any convenient location in the neutron-reflecting coolant loop to assist or drive the circulation of the liquid neutron-reflecting coolant. In the embodiment shown, the fuel is shell-side and the reflector material which is also the coolant is tube-side. The shell and tubes are made of some structural material that is solid at the operating temperatures. The circulation of fuel salt (illustrated by a dashed line 2406) from the reactor core 2404 into and through the shell side of the primary heat exchanger 2410 and back into the bottom of the reactor core 2404. Baffles 2412 are provided in the shell to force the fuel salt to follow a circuitous path around the tubes of the heat exchanger. Reflector/coolant flows through the tube-side of the heat exchanger 2410, but before entering the bottom of the heat exchanger first flows down the length of a coolant inlet channel adjacent to the sides and bottom of the containment vessel 2418. In an embodiment, a solid layer of reflector material may form on the inner surface of the containment vessel, especially if the exterior of containment vessel 2418 is cooled. This is acceptable as long as it does not interfere with the flow of the reflector/coolant. The reflector/coolant then enters the tubes of the heat exchanger by flowing through the lower tube sheet 2431, which is illustrated as being level with the bottom of the reactor core 2404. The reflector/coolant exits the tubes of the heat exchanger at the upper tube sheet 2432, which is located in FIG. 24 some distance above the reactor core 2404 and containment vessel 2418. FIG. 24 illustrates a region 2434 within the shell of the heat exchanger that is above the level of fuel salt in the reactor core 2404. This region may be filled, except for the penetrating tubes, with any reflecting or moderating material, for example filled with a different or the same reflector material as the reflector/coolant. In FIG. 24, the upper reflector 2408A and lower reflector 2408B are illustrated as distinct from the circulating reflector/coolant material. In an alternative embodiment, the upper reflector 2408A, lower reflector 2408B, and inner reflector 2408C may all be in fluid communication as shown in FIGS. 22 and 23. For example, reflector material may be routed into reactor 2400 along the interior surface of the side of the containment vessel 2418, as shown, but then routed along the bottom of the containment vessel 2418 before rising and making a U-turn to flow adjacent to the bottom of the reactor core 2404, as shown in FIG. 23. The reflector material may also be routed into the upper reflector 2308A from which it can be removed at a central location, also as shown in FIG. 23. A pump (not shown), or at least the impeller of a pump, may be provided to assist in fuel salt circulation or reflector/coolant circulation. For example, an impeller may be provided in one or both of the heated fuel salt inlet to the primary heat exchanger at the top of the reactor core 2404 or (as discussed in greater detail below) the cooled fuel outlet of the shell of the primary heat exchanger at the bottom of the reactor core 2404. In yet another embodiment, reflective coolant may be flowed through upper and lower axial reflectors to advect away any heat generated in these reflectors in a circulation loop that is separate from the primary cooling loop. In yet another embodiment of a reflector design, a ‘breed and burn blanket’ may be provided surrounding the main core. In this embodiment, a reflector ‘blanket’ containing uranium could be provided, either as the only reflector or as a second reflector located inside (between the core and the primary reflector) or outside of the primary reflector. The uranium in the reflector could be either liquid or solid, and could be uranium metal, a uranium oxide, a uranium salt or any other uranium compound. The uranium in the reflector will reflect neutrons but will also breed plutonium over time, thus becoming a source of fuel. It will be clear that the systems and methods described herein are well adapted to attain the ends and advantages mentioned as well as those inherent therein. Those skilled in the art will recognize that the methods and systems within this specification may be implemented in many manners and as such is not to be limited by the foregoing exemplified embodiments and examples. In this regard, any number of the features of the different embodiments described herein may be combined into one single embodiment and alternate embodiments having fewer than or more than all of the features herein described are possible. While various embodiments have been described for purposes of this disclosure, various changes and modifications may be made which are well within the scope contemplated by the present disclosure. Numerous other changes may be made which will readily suggest themselves to those skilled in the art and which are encompassed in the spirit of the disclosure.
048428094	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The invention is related to storage of nuclear fuel rods and in particular to a rod arraying system which consolidates two fuel assemblies into the rack space of one standard fuel assembly. 2. General Background Storage space at nuclear reactor sites is very limited. This results in a need to be able to consolidate nuclear fuel being stored. A consolidation ratio of 2 to 1 can be obtained by disassembling two fuel assemblies and repackaging the fuel into one canister which will fit into the same rack space as a standard fuel assembly. Since there is seldom enough space at a utility to allow horizontal handling of fuel, the majority of existing designs and equipment directed toward fuel consolidation handle fuel in a vertical orientation. In most of these systems fuel rods are pulled or pushed from a fuel assembly in groups or bundles and put into a funnel which reconfigures the fuel rod bundle. Bundles may contain a full complement of rods for a complete fuel assembly or rods from a partial assembly. The funnel is a long device that accepts a rod bundle at one end in the same configuration as it is taken from a fuel assembly (rectangular pitch) and gradually changes the configuration of the bundle into a close packed triangular pitch rod configuration at the other end. The reconfigured rods are then fed into a canister for storage. This approach to fuel rod consolidation has met with limited success since it does not lend itself to quick recovery from an off normal situation such as a broken fuel rod. The funnel approach does not lend itself to other fuel related operations such as burnable poison rod consolidation and does not support fuel reconstitution, eddy current inspection, and recaging of damage fuel assemblies. It can be seen that a more versatile concept is needed in this area. SUMMARY OF THE INVENTION The present invention solves the above problem in a straightforward manner. What is provided is a rod arraying device which uses the principle of single rod transfer. The handling of one rod at a time affords the most operational control and eliminates the chance of major portions of fuel assemblies getting hung up as a result of equipment failures, power failures, or rod breakage. The rod araying device works in conjunction with the consolidation canister. The rod arraying device has two scalloped plates with identical edge profiles. The plates are laid one on top of the other and offset in the front-to-back and side-to-side directions. The relative position of the plates is maintained by bearing guides which allow forward and backward cycling of the bottom guide by an air cylinder. The rod loading pattern provides gaps between rods that allow the rod arraying devices to hold each rod in position.
	summary
053316762	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to the field of temperature measurement and control, in particular in an induction furnace for heating the end of a zirconium alloy nuclear fuel rod in the presence of an oxidizing gas, to form a protective zirconium oxide layer in an area of the fuel rod at which a fuel assembly structure tends to fret captured debris against the fuel rod. According to the invention, a probe that is configured to correspond to a fuel rod is provided with an array of spaced thermocouples for monitoring the temperature contour obtained in the induction furnace, enabling precise measurement and control of the temperature profile that occurs when a fuel rod is treated in the induction furnace under the same conditions. 2. Prior Art In a nuclear power plant, coolant is heated in a vessel disposed along a primary coolant circuit, by nuclear fuel in the form of vertically elongated fuel rods carried in fuel assemblies. Each fuel rod comprises a stack of enriched uranium pellets in a hollow tube typically made of a zirconium alloy such as Zircalloy. The zirconium alloy presents a low nuclear cross section to neutrons that carry on the fission reaction leading to heating of the fuel. Zirconium is a highly active metal that, like aluminum, seems passive because a stable and cohesive thin zirconium oxide (ZrO.sub.2) film forms on the surface in the presence of oxygen, e.g., when exposed to air or water. Formation of the oxide layer is accelerated with heating. Zirconium is advantageous for its subatomic properties, but is not the most durable of metals. It is possible to treat zirconium fuel rod tubes prior to their use in a reactor, to add a protective cladding layer that is more durable, thicker and/or more chemically resistant than the basic tube material. Such a cladding can comprise a plated-on alloy or a distinct metal. Heating of the fuel rod tube may be involved in the process. Typically the entire tube is treated; however, it is also possible to treat only particular areas that are considered vulnerable to corrosion or the like, for example the inside of the tube. Particular claddings are chosen for corrosion resistance, especially resistance to corrosion from reaction of the zirconium with iodine and other elements released during fission of the nuclear fuel. Examples of protective treatment of the inside or outside of fuel rod tubes are disclosed, for example, in the following patents: ______________________________________ 4,100,020 Andrews 4,233,086 Vesterlund 4,411,861 Steinberg 4,609,524 Ferrari 4,613,479 Foster 4,659,545 Ferrari 4,675,153 Boyle et al. 4,894,203 Adamson 5,026,516 Taylor 5,073,336 Taylor 5,137,683 Hertz et al. ______________________________________ In pressurized water and boiling water reactors, the fuel rods are grouped in fuel assembly structures that have vertically spaced grids for holding the fuel rods in a parallel array, whereby a number of the fuel rods can be handled as a fuel assembly unit. The liquid coolant is heated in a vessel by fission in the fuel rods, causing a vigorous and turbulent upward flow of coolant over the fuel rods due to convection. The fuel assembly grids are openwork panels disposed perpendicular to the elongation of the fuel rods, with spring structures that bear against the fuel rods to hold them in place. However, in the turbulent flow of the reactor coolant, loose metallic debris may be stopped by the grids, especially at the endmost grid facing the direction of flow. This captured debris vibrates against the fuel rods, leading to fretting damage to the zirconium alloy tube in the area of the endmost grid. A breach of the fuel rod tube can lead to release of fuel into the coolant, which is undesirable due to the resulting circulation of radioactive material with the coolant. Fretting damage to the fuel rods has been found to occur primarily during the first cycle of their irradiation. Fissile heating of the fuel rods in the water coolant of the reactor during use thickens the zirconium oxide layer on the outer surfaces of the fuel rods and thereafter protects the fuel rods from fretting damage. Operational temperatures and pressures in a pressurized water reactor, for example, may be on the order of 300.degree. to 400.degree. C. and 150 bar. As described in commonly owned U.S. patent application Ser. No. 08/025,361, filed Mar. 2, 1993, it is possible to pretreat the fuel rod tubes to form a protective zirconium oxide layer prior to installation of the fuel rods. The protective layer is formed along the endmost four to eight inches (10 to 20 cm) of the fuel rod and reduces or eliminates fuel rod failure during initial use. This end portion of the fuel rod is the approximate length that protrudes from the lowermost grid of the fuel assembly, in the direction facing the coolant flow, i.e., the area in which the lowermost grid is likely to capture debris that will fret against the fuel rod. The protective zirconium oxide layer can be formed by heating the fuel rod tube in the presence of oxygen. The thickness of the resulting ZrO.sub.2 layer is a function of the time of heat treatment, the temperature, and the oxygen concentration in the treating atmosphere, typically air. It is desirable to form a uniform coating that completely encompasses the end of the tube, and has a depth of two to fifteen microns. Formation of a protective oxide by heating in this manner is of course much easier than application of an alloy cladding, for example requiring a plasma arc or other process, such as in U.S. Pat. No. 5,227,129 - Bryan et al. There are a variety of means by which a fuel rod tube can be heated, for example using convection, laser irradiation, application of a flame, etc. An advantageous method is heating via electrical induction. The end of the fuel rod tube is placed in an electrical induction furnace having coils coupled to an AC power source, for inducing a current in the metallic zirconium material. Induced eddy currents dissipate electrical power by resistance heating. This form of heating is advantageous in that the power can be concentrated at the area to be treated (at least subject to conduction of the heat to the remainder of the fuel rod tube). Preferably, the induction furnace is only barely larger than the fuel rod or rods being treated. In this manner, the electromagnetic field intensity can be maximized by minimizing the gap between the coils (or the ferromagnetic material coupling the field to the fuel rod) in the magnetic circuit. To achieve uniform application of electromagnetic energy, a series of adjacent or spaced coil pairs can be disposed on opposite sides of the fuel rod tube, each of the coils in a pair being energized at opposite polarity and each coil pair encompassing a limited axial length of the fuel rod. These individual coil pairs can be controlled separately, for applying the precise power level needed to achieve uniform heating of the fuel rod tube. It is advantageous to apply the minimum power necessary to obtain the required thickness of ZrO.sub.2 over all the end of the fuel rod tube to be treated. A uniform coating of ZrO.sub.2 that is of sufficient thickness over all the area of treatment requires precise temperature control. To achieve the uniform heating needed over the length of the fuel rod end, some means is needed to measure the induced heat from the respective coils, and to adjust or control the power level applied to the coils as necessary. It is not entirely adequate simply to apply equal power levels to each of the coil pairs, because the effects of the induction heating may be uneven even if the field strengths are equal, due to the variations in the eddy currents induced in the fuel rods occurring due to differences in geometry along the fuel rods. For example, the induced currents may vary between the center of the treated length of the fuel rod and the extreme end of the fuel rod, or between the center and the area at which the treated portion meets the proximal portion of the fuel rod, due to end effects and due to the adjacent conductive metal, respectively. The present invention concerns a method and apparatus for sensing the temperature effects of electromagnetic induction in a fuel rod by providing a temperature probe structured to simulate the conductive and/or resistive aspects of the fuel rod tube, and has an array of thermocouples disposed to measure the temperatures at specific spaced points. In this manner a temperature profile can be measured and used to adjust the electromagnetic field strengths of the coil pairs, for obtaining a desired temperature profile when the probe is replaced by an actual fuel rod to be treated. Temperature probes for measuring the heat in a furnace are known generally. In a typical application, the probe is moved to different areas of the furnace in order to develop a temperature profile from a series of successive measurements. It is also possible to use an array of temperature sensors on a probe, for example as shown by U.S. Pat. Nos. 4,176,554 or 4,242,907, both to Kazmierowicz, or 4,098,122- Landman et al. For such uses, the spaced temperature sensing elements of the probe are intended to measure the ambient temperature at different points in the furnace or kiln. It is assumed that the ambient temperature as so measured is the temperature to which the workpieces will be heated when at the corresponding location. The furnace or kiln is then adjusted to obtain a desired temperature profile. This technique cannot be used effectively if the presence or movement of the workpiece being heated in the furnace has an effect on the temperature to which the workpiece is heated. In a tunnel furnace, for example, cool workpieces entering the furnace reduce the ambient temperature there, other things being equal. Similarly, where a tunnel furnace has workpieces moving from a zone at one temperature into a zone at another temperature, it will take a certain time for the workpieces to reach the temperature of the new zone, assuming that the workpieces remain long enough to reach the ambient temperature at all. The reason for providing a probe having spaced temperature sensors is to obtain a measurement of the differences in temperature from point to point in the furnace, for example due to the workpieces, so that such effects can be addressed. Temperature probes as disclosed in Kazmierowicz or Landman et al. are not suitable for measuring the temperature profile of an electromagnetic induction furnace, for adjustment of the field generating means as needed to develop the required temperature profile in workpieces when inserted to be heated. The electrical induction form of heating apparatus is particularly affected by the presence of the workpiece, because for the most part the heat is generated in the workpiece rather than in the ambient air of the furnace. Although indirect temperature sensing means could be installed in the furnace to sense the workpiece temperature at different points, this is a complex and expensive solution to the problem. According to the present invention, a probe having spaced temperature sensors is configured to resemble a workpiece, in particular the end of a fuel rod tube to be heated for producing a protective oxide layer, whereby currents induced in the probe and the temperature to which the probe is heated, closely model the situation for actual fuel rod workpieces. SUMMARY OF THE INVENTION It is an object of the invention to provide a temperature probe for monitoring the temperature profile obtained in conductive parts heated in an electromagnetic induction furnace. It is another object of the invention to simplify the measurement of temperature in the heat treating of nuclear fuel rod tubes by electromagnetic induction. It is a further object of the invention to enable precise temperature control in a compact induction furnace, sized to closely complement the parts being heated. These and other objects are accomplished by heating nuclear fuel rod tubes of zirconium alloy in an induction furnace to produce a protective oxide coating two to fifteen microns in thickness. The furnace cavity is only slightly larger than the tubes and receives the endmost eight or so inches of the tube. The furnace is controllable in zones along the tube, for example in one inch increments. To calibrate the furnace to produce the desired temperature profile, typically a flat profile at a temperature between 650.degree. and 750.degree. C..+-.1.5.degree. C., a temperature calibration probe is provided with spaced thermocouples for sensing the temperature developed in the probe at each of the zones when heated. The probe preferably is made of Inconel 600 alloy or the like, and is dimensioned and shaped to correspond closely to the dimensions of the fuel rod tubes, including having a closed chamfered end. At the opposite end the probe protrudes from the furnace, where the thermocouple leads are terminated. The leads pass through a potting compound in the probe, such as magnesium oxide. Whereas the probe conductive structures are substantially identical to corresponding structures of the tube with respect to heating by electromagnetic induction, the probe responds to the electromagnetic field in the induction furnace substantially the same as does the end of the tube, permitting calibration of the induction furnace zones for a desired temperature profile, e.g., a flat profile along the length of the tube, notwithstanding differences in induced currents that would otherwise occur due to the end of the tube or the adjacent tube material.
053944470	claims	1. In a nuclear power plant having a nuclear steam supply system including a multiplicity of components which operate together to perform plant functions including critical safety functions, which must be accomplished to keep the plant in a safe, stable condition whereby the health and safety of the public is preserved; means for measuring plant operating variables and for generating operating parameter signals from said measured variables; a control room including data processing means responsive to the operating parameter signals for displaying monitoring information including parameter values and parameter alarms to the operator; the improvement in the means for displaying monitoring information about the critical safety functions comprising: means for displaying on a large screen situated for visibility throughout the control room, storing a hierarchy of display pages including continuously determining the status of the critical functions by comparing parameter signals against plant critical function acceptance criteria; highlighting a particular critical function descriptor on the apex page when the acceptance criteria for the particular critical function is not satisfied; and presenting to the operator means for accessing the first, second, and third level display pages whereby the operator can diagnose the success paths relating to the highlighted critical function descriptor. at least one key parameter representing each critical function and an alarm descriptor with the key parameter display when a parameter alarm associated with the key parameter arises, and the status of at least one success path for each critical function, including the state and controllability of the associated component configuration, and an alarm descriptor with the status when the success path is unavailable or underperforming. the shape coding is hollow/solid to indicate active/inactive status, and the color coding is green/red to indicate (closed or off/open or on) and yellow to indicate a parameter alarm. continuously displaying said apex display, on a large, centrally located display screen visible throughout the control room, and accessing said first, second and third level displays on at least one operator's panel including another display screen. the status, availability and performance information is displayed on the apex display, and the display screen on the operator's panel can display the apex display. continuously determining the status of the critical function success paths including the availability, current operating state, and current performance of critical function success paths and delivering the results for storage in said means for storing; generating a success path unavailability alarm if a success path cannot be actuated to achieve minimum acceptable performance criteria; and generating a success path performance alarm if a success path in actuated but the quality of performance is below a minimum acceptance criteria. 2. The improvement of claim 1, wherein the plant functions include critical mission functions which must be accomplished in order to ensure uninterrupted power generation in the plant and the means for displaying displays each of said descriptors, status indicators, key parameter alarms, and unavailability alarms for both the safety and mission critical functions. 3. The improvement of claim 1, wherein the plant can be operated in any one of a plurality of modes and the critical functions and success paths are dependant on the operating mode. 4. The improvement of claim 3, wherein the modes include normal operation, heatup/cooldown, cold shutdown/refueling, and post-trip. 5. The improvement of claim 1, wherein the means for displaying on a large screen also displays a success path performance alarm if a success path has been actuated but the quality of performance is below a minimum acceptance criteria. 6. In a nuclear power plant having a nuclear steam supply system including a multiplicity of components which operate together to perform critical plant functions; means for measuring plant operating variables and for generating operating parameter signals from said measured variables; a control room including data processing means responsive to the operating parameter signals for displaying monitoring information including parameter values and parameter alarms to the operator; means responsive to the operating parameter signals for automatically initiating operation of safety related components upon the occurrence of certain abnormal events, means by which the operator can control safety related components, and means by which the operator can control components relating to power generation in the plant; the improved method of data processing to display monitoring information about the critical plant functions, comprising: 7. The improvement of claim 6, including displaying on the apex page, 8. The improvement of claim 7, wherein the status display of each success path on the apex page comprises a geometric figure with shape, texture, and color coding. 9. The improvement of claim 7, wherein 10. The improvement of claim 6, wherein each success path descriptor on the directory is touch sensitive such that by touching any one of such descriptors, the operator can retrieve the second level display page associated with said touched success path descriptor. 11. The improvement of claim 6, wherein the plant can be operated in any of a plurality of modes and the critical function acceptance criteria is dependant on the plant operating mode including power production and post trip modes. 12. The improvement in claim 6, wherein each of the second level displays includes a time trend of the key parameter of the associated critical function. 13. The improvement in claim 6, wherein each of the second level displays includes a high level mimic diagram of the success paths for the associated critical function. 14. The improvement of claim 7, including 15. The improvement of claim 14, wherein 16. The improvement of claim 6, including
	claims	1. An end support system for a container for shipping nuclear fuel, comprising: first and second elongated metal plates generally parallel to and spaced from one another and third and fourth metal plates generally parallel to and spaced from one another, said third plate being secured at opposite ends to ends of said first and second plates and said fourth plate being secured at opposite ends to opposite ends of said first and second plates thereby forming a generally rectilinear metal end frame; an elongated metal cross-plate secured at opposite ends to said third and fourth plates, respectively, at locations intermediate ends of said third and fourth plates, said cross-plate extending generally parallel to said first and second plates and lying generally in a plane defined by said metal end frame; a metal reinforcing member secured to said cross-plate and projecting from one side thereof and generally out of said plane, said member being located intermediate said opposite ends of said cross-plate and inwardly of said third and fourth plates; and at least a pair of metal supports connected to said metal end frame and extending generally perpendicular to said end frame along opposite sides of said metal end frame for securement to the sides of said container. 2. A system according to claim 1 wherein said reinforcing member and said supports project in the same direction from the plane of said metal end frame. claim 1 3. A system according to claim 1 wherein said reinforcing member is elongated and extends along a central region of said crosspiece, terminating at opposite ends short of the ends of said crosspiece. claim 1 4. A system according to claim 3 wherein said reinforcing member comprises a channel. claim 3 5. A system according to claim 1 wherein said metal supports comprise a pair thereof secured to each of opposite sides of said metal end frame. claim 1 6. A system according to claim 1 wherein said metal supports comprise four arms secured to said end frame adjacent corners thereof. claim 1 7. A container for shipping nuclear fuel, comprising: an elongated container body having sides, a top and bottom and opposite ends, a discrete metal and frame for reinforcing each of said opposite ends of said container body, each said metal end frame comprising first and second elongated metal plates generally parallel to and spaced from one another and third and fourth metal plates generally parallel to and spaced from one another, said third plate being secured at opposite ends to ends of said first and second plates and said fourth plate being secured at opposite ends to opposite ends of said first and second plates thereby forming said metal end frame in a generally rectilinear configuration; an elongated metal cross-plate secured at opposite ends to said third and fourth plates, respectively, and at locations intermediate ends of said third and fourth plates, said cross-plate extending generally parallel to said first and second plates and lying generally in a plane defined by said metal end frame; a metal reinforcing member secured to said cross-plate and projecting from one side thereof and generally out of said plane, said member being located intermediate said opposite ends of said cross-plate and inwardly of said third and fourth plates; at least a pair of supports connected to said metal end frame and extending generally perpendicular to said end frame along opposite sides of said metal end frame for securement to said container; said metal end frames being secured to said container body ends respectively, with said supports straddling opposite sides of said container body ends. 8. A container according to claim 7 wherein said ends and said sides of said container body are formed at least in part of wooden framing elements, each said metal end frame and said supports thereof being secured to said wooden framing elements along said ends and said sides, respectively, of said container body. claim 7 9. A container according to claim 7 wherein each said container body end includes wooden framing elements underlying said metal plates, respectively, and in a plane parallel to said plane of said end frames, said reinforcing member extending between a pair of said wooden frame elements. claim 7 10. A container according to claim 7 wherein each said container body end includes wooden end framing elements forming a wooden end frame underlying said metal end plates, respectively, and in a plane parallel to said plane of said metal end frame, an end panel on each said container body end secured to a side of said wooden end framing elements opposite said metal end frame, said reinforcing member projecting from the plane of said metal end frame to reinforce said end panel. claim 7 11. A container according to claim 10 wherein said reinforcing member extends between a pair of said wooden frame elements of said wooden end frame and comprises a channel. claim 10 12. A container according to claim 7 wherein said metal supports comprise four arms secured to said end frame adjacent corners thereof. claim 7 13. A container for shipping nuclear fuel, comprising: an elongated container body having sides, a top and bottom and opposite ends; a metal end frame for overlying and reinforcing each of said opposite ends of said container body and lying in a plane, said ends and said sides of said container body being formed at least in part of wooden framing elements; a reinforcing member secured to said metal end frame and projecting from one side of said metal end frame and generally out of said plane, said member being located intermediate opposite edges of said metal end frame; at least a pair of metal supports connected to said metal end frame and extending generally perpendicular to said metal end frame along opposite sides of said metal end frame for securement to said container; each said metal end frame and said supports being secured to said wood framing elements along said ends and sides of said container body, respectively, with said supports straddling opposite sides of said container body ends, said reinforcing member engaging at least a portion of said container body to reinforce the container body. 14. A container according to claim 13 wherein said metal end of frames are discrete and spaced from one another at said opposite ends of said container body, said container body including wooden end framing elements forming wooden end frames underlying said metal end plates, respectively, said wooden end frames lying in planes parallel to planes containing said metal end frames, an end panel on each said container body end secured to a side of said wooden end framing elements opposite said metal end frame, said reinforcing member projecting form the plane of said metal end frame to reinforce said end panel. claim 13 15. A container according to claim 14 wherein said reinforcing member extends between a pair of said wooden end frame elements and comprises a channel. claim 14 16. A container according to claim 15 wherein said metal supports comprise four arms secured to said end frame adjacent corners thereof. claim 15
	summary
	abstract	The invention relates to nuclear engineering and can be used for fuel clusters of nuclear reactors, for distancing and fixing fuel elements, in particular in the fuel clusters of PWR and BWR reactors. The inventive distance lattice comprises cells which are used for mounting the fuel elements or guide channels and are formed by perpendicular crossing plates. Bent blades for mixing a coolant are embodied on the plate edges at the output of said coolant. Each cell is provided with an insertable distancing element for fixing the fuel element. Said invention makes it possible to increase the turning rigidity of the cells and the stability thereof, to simultaneously reduce the size of the fixation of the fuel elements or the guiding channels in the cells and to decrease the hydraulic resistance of the lattice.
048662810	claims	1. An irradiation plant comprising an irradiation chamber (1) and a conveyor system for conveyor units (5) bearing articles for irradiation and moved past a radiation source (2) the system comprising an even number of irradiation tracks (3) disposed symmetrically to the radiation source (2) and extending between a first transverse track (4) and a second track (4A) each connecting the irradiation tracks (3) at a respective end, and one transverse track (4) being connected to the entry section (10, 15) and the exit section (16, 11) of the conveyor system to and from the irradiation chamber (1), characterised in that each transverse track (4, 4A) comprises a shift device (8, 8A) for loading, unloading and changing over the conveyor units (5) on the irradiation tracks (3) and the conveyor units (5) have control elements (70) for presetting the path of the conveyor unit (5) in the conveyor system and acting on sensors (71) of a conveyor-system control means, and the control means is constructed so that whenever a conveyor unit (5) is loaded on to or unloaded from an irradiation track (3), a shift device (8, 8A) is disposed on each side of the conveyor unit (5), and at the end of each loading or unloading process, not more than one of the two facing shift devices (8, 8A) is loaded with a conveyor unit (5). 2. An irradiation plant according to claim 1, characterised in that the control elements are engageable and disengageable levers (70) secured to the conveyor units (5) and the levers act only in one of these two positions on the sensors (71) disposed along the conveyor paths. 3. An irradiation plant according to claim 1 or 2, characterised in that the radiation source (22) is in the form of a wall, guided by a hoisting device (21) on rails (23), and lowerable into a water container (20). 4. An irradiation plant according to claim 3, characterised in that rack or chain (28) is disposed along each rail (23) and engages a pinion (27) disposed on the radiation source (22). 5. An irradiation plant according to claim 1 characterised in that the radiation source comprises at least two wall-shaped radiators (22). 6. An irradiation plant according to claim 1 characterised in that conveyor units (5) comprise a number of superposed conveyor surfaces (57). 7. An irradiation plant according to claim 6, characterised in that a change station (17) is provided in the irradiation chamber (1) for changing the articles for irradiation on the superposed conveyor surfaces (57). 8. An irradiation plant according to claim 1 characterised in that the conveyor units (5) on each conveyor surface (57) have a releasable securing brake (60), the conveyor surfaces (57) have rollers (56) supporting the bearers (6, 58) of the articles for irradiation, and the bearers (6, 58) are loaded and unloaded on the rollers (56) by co-operating pressure-medium actuated slides (45). 9. An irradiation plant according to claim 1 characterised in that drive chains (55) with cams (79) are available for driving the conveyor units (5) and engage in grooves (55) of coupling members (53) of the conveyor units (5). 10. An irradiation plant according to claim 1 characterised in that each of the two transverse tracks (4, 4A) has a single shift device (8, 8A) and the two shift devices (8, 8A) are driven in the same direction on the transverse tracks (4, 4A). 11. An irradiation unit according to claim 10, characterised in that each shift device (8, 8A) has a drive (7) for the conveyor units (5) and the drive (7) is coupled to the drive (8) of the shift devices (5) so that only one device (5) is operative at a time. 12. An irradiation plant according to claim 1 characterised in that at least one parking track (31) for receiving conveyor units (5) is in the irradiation chamber (1). 13. An irradiation plant according to claim 1 characterised in that the tracks (3, 3, 8, 8A, 10, 11, 15, 16, 31) of the conveyor system are constructed as suspended rail tracks. 14. An irradiation plant accordint to claim 1 characterised in that the radiation source (2, 22) is a radioactive radiation source.
	abstract	As the operator sets a direction of moving a turn table, a view field FOV (z) of an X-ray detector in the rotation axis direction of the turn table, and the number of imaging times n and gives a continuous imaging command, the imaging operation of collecting X-ray fluoroscopic data while rotating the turn table and the operation of moving the turn table in the setup direction by FOV (z) are repeated and imaging is conducted n times. Then, the provided data is reconstructed to provide tomograms. It is possible to provide a continuous tomogram even if a high imaging magnification is set to provide high-resolution three-dimensional data and a necessary region for the three-dimensional data cannot be covered in the field of view of the X-ray detector.
050930737	description	EXAMPLE 1 The samples a) of ferritic chromium steel were treated at room temperature (290 K. to 295 K.) for 16 hours with a solution of 0.05 mol each of chromic acid and permanganic acid. After intermediate rinsing, a decontamination factor (ratio of measured activity before and after the treatment) of 2 was found. A further treatment at room temperature in an aqueous 0.1 mol solution of oxalic acid under the action of ultrasonics led to a decontamination factor of about 20 after 15 minutes and to a decontamination factor of more than 100 after 6 hours. After the treatment, the decontaminated surfaces of the samples were metallically bright and not noticeably attacked either macroscopically or microscopically. EXAMPLE 2 Samples c) of nickel/chromium/iron alloys of trade name INCONEL 600 were treated at room temperature for 16 hours with a solution of 0.1 mol of chromic acid and 0.004 mol of potassium permanganate. After intermediate rinsing, a decontamination factor of only 1.2 was found. After a further treatment at room temperature with an aqueous solution of 0.1 mol of oxalic acid for 6 hours under the action of ultrasonics, a decontamination factor of 12 was determined. EXAMPLE 3 Samples a) of ferritic chromium steel, samples b) of austenitic stainless steels and samples c) of INCOLOY 800 and of INCONEL 600 were each treated for 16 hours at room temperature in aqueous solutions with 0.01 to 0.1 mol of chromic acid and 0.001 to 0.05 mol of permanganic acid, the chromic acid/permanganic acid ratio being between 1:10 and 25:1. The samples were then each further treated for 6 hours at room temperature in an aqueous solution of 0.1 mol of oxalic acid under the action of ultrasonics. Finally, decontamination factors of between 10 and 1000 were measured on all the samples, depending on the oxidative treatment and on the sample material. EXAMPLE 4 Samples a) of ferritic chromium steel and samples c) of INCONEL 600 were each treated for 16 hours at room temperature in a solution of 0.1 mol of chromic acid and 0.05 mol of permanganic acid. After a subsequent treatment with a water jet of 2.4 kbar (240 Pa) pressure at a treatment rate of 3.6 m.sup.2 /hour, decontamination factors of about 30 were measured on the samples a) of ferritic chromium steel, and decontamination factors of more than 100 on the samples c) of INCONEL 600. Extensive further investigations showed that the surfaces of the base materials were not attacked by these treatments. EXAMPLE 5 Samples c) of INCONEL 600 were sprayed for 16 hours at room temperature with a solution of 0.05 mol of chromic acid and 0.002 mol of permanganic acid. After a subsequent further treatment with a water jet as in Example 4, decontamination factors of between 20 and 80 were determined. EXAMPLE 6 A paste was prepared from an aqueous solution of 0.4 mol of chromic acid and 0.1 mol of permanganic acid by addition of a thickener which is available on the market under the trade name AEROSIL (registered trademark of Degussa). This paste was spread on the contaminated surfaces of samples a) of ferritic chromium steel. After a period of action of 16 hours, the samples were treated with a water jet as in Example 4. The resulting decontamination factors were between 5 and 15. The tests described by way of example and further extensive investigations showed that the materials normally used in reactor construction for the cooling circuits are not damaged by the treatments using the process according to the invention, irrespective of whether the components decontaminated in this way have aged or have been heat-treated, welded or deformed.
	summary
	claims	1. A garnet-type crystal for a scintillator that is represented by General Formula (1), (2), or (3),Gd3-x-yCexREyAl5-zGazO12 (1)wherein in Formula (1), 0.0001≦x≦0.15, 0≦y≦0.1, 2.5≦z≦3.5, and RE represents at least one selected from Y and Yb,Gd3-a-bCeaLubAl5-cGacO12 (2)wherein in Formula (2), 0.0001≦a≦0.15, 0.1<b≦3, and 3<c≦4.5,Gd3-p-qCepRE′qAl5-rGarO12 (3)wherein in Formula (3), 0.0001≦p≦0.15, 0.1<q≦3, 1<r≦4.5, and RE′ represents Y or Yb. 2. The garnet-type crystal for a scintillator according to claim 1,wherein a fluorescence component has a fluorescence lifetime of not longer than 100 ns. 3. The garnet-type crystal for a scintillator according to claim 1,wherein the intensity of a long-life fluorescence component having a fluorescence lifetime exceeding 100 ns is not higher than 20% of the intensity of the entire fluorescence components. 4. The garnet-type crystal for a scintillator according to claim 1,wherein a peak emission wavelength of the fluorescence component is equal to or longer than 460 nm and equal to or shorter than 700 nm. 5. The garnet-type crystal for a scintillator according to claim 1,wherein an amount of luminescence is 20,000 photons/MeV or more. 6. A radiation detector comprising:a scintillator constituted with the garnet-type crystal for a scintillator according to claim 1; anda light receiver that detects luminescence from the scintillator. 7. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (1). 8. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (1) and in the Formula (1), RE represents Y. 9. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (1) and in the Formula (1), RE represents Yb. 10. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (1) and in the Formula (1), 0.003≦x≦0.15. 11. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (2). 12. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (2) and in the Formula (2), 0.015≦a≦0.09. 13. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (2) and in the Formula (2), 0.015≦a≦0.09 and 3<c≦4 . . . 0. 14. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (3). 15. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (3) and in the Formula (3), RE′ represents Y. 16. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (3) and in the Formula (3), RE′ represents Yb. 17. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (3) and in the Formula (3), 0.015≦p≦0.09. 18. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (3) and in the Formula (3), 3<r≦4.5. 19. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (3) and in the Formula (3), 0.5≦q≦3 and 2≦r≦4. 20. The garnet-type crystal for a scintillator according to claim 1,wherein the garnet-type crystal is represented by the General Formula (3) and in the Formula (3), 0.5≦q≦1.5 and 2.5≦r≦3.5.
	abstract	There is disclosed an apparatus applied to exposure by a charged beam, having a pattern information acquiring section acquiring information on a character projection pattern formed in a character projection aperture mask, a first information storing section storing information on a reference pattern, and an identifying section identifying a shape of the character projection pattern by comparing the information on the character projection pattern with the information on the reference pattern.
	summary
06295332&	abstract	The present invention is directed to the production of high quality semi-conductor devices created at speeds and in sizes that far exceed current x-ray lithography capabilities. The steps involved in the method include the use and development of horizontal beams from a synchrotron or point source of x-ray beams; preparation of submicrometer, transverse horizontal and vertical stepper stages and frames; providing a stepper base frame for the proper housing and mating of the x-ray beam; minimizing the effects of temperature and airflow control by means of a environmental chamber; transporting, handling and prealigning wafers and other similar items for tight process control; improving the control and sensing of positional accuracy through the use of differential variable reluctance transducers; controlling the continuous gap and all six degrees of freedom of the wafer being treated with a multiple variable stage control; incorporating alignment systems using unambiguous targets to provide data to align one level to the next level; beam transport, shaping or shaping devices, to include x-ray point sources; using an inline collimator or concentrator for collimating or concentrating the x-ray beams; and, imaging the mask pattern at the precise moment for optimum effectiveness.
043199593	abstract	In supervising the channel stability of a nuclear reactor, the thermal power and the coolant flow quantity in each fuel assembly are determined by signals produced by a plurality of neutron flux detectors installed in the nuclear reactor, and signals regarding other operating conditions of the reactor, and the thermo-hydrodynamic stability of each fuel assembly is judged by the above-mentioned thermal power and coolant flow quantity as well as the measured values of such parameters as inlet subcooling of the core and of the pressure in the reactor vessel. Or the stability limit of a selected one of the parameters is determined. Then, the threshold of the thermo-hydrodynamic stability or the stability limit is compared with the actually measured value thereof so as to determine a stability margin.
048809896	abstract	A radioaerosol delivery apparatus particularly adapted for the subsequent disposal of radioactively contaminated elements is described. The apparatus includes a shielding container including an outer shell and an inner shell supported within the outer shell. The inner shell is formed with an inner wall and an outer wall defining a space therebetween for receiving radiation shielding material. The inner wall is formed to provide a surface conforming substantially to the contour of the radioaerosol source and a surface conforming substantially to the contour of a portion of a transport manifold supported thereon. A removable cover including radiating shielding material is formed with a portion conforming generally to the contour of the transport manifold. The inner wall of the container and the cover define at least one opening therebetween to permit fluid communication from the transport manifold to a patient utilizing the apparatus.
	abstract	A module including a casing extending in a longitudinal direction, a bundle of fuel rods encased in and supported by the casing and connector provided on the casing for connecting the casing side-by-side to the casing of at least one other module to obtain a nuclear fuel assembly having a channel box defined by the casings of the assembled modules and of larger cross-section than the casing of each of the assembled modules and a bundle of fuel rods of larger cross-section than that of each the assembled modules.
045377103	claims	1. A method of fixing metal ions contained in liquid low and intermediate level nuclear wastes for long-term storage comprising: contacting the liquid waste with an ion-exchange material which is a modified tobermorite containing aluminum isomorphously substituted for silicon and containing per 100 grams, about 1 to 200 milliequivalents of an alkali metal selected from the group consisting of sodium, potassium and mixtures thereof, whereby the metal ions in the liquid waste are taken up by the ion-exchange material; separating the ion-exchange material from the liquid; mixing the ion-exchange material containing the ions with portland cement to form a mixture containing up to about 40 weight percent ion-exchange material; and solidifying the mixture, thereby fixing the ions for long-term storage. contacting a solution containing cesium and other metal ions with a modified-tobermorite containing aluminum isomorphously substituted for silicon and containing per 100 grams, about 1 to 200 milliequivalents of an alkali metal selected from the group consisting of sodium, potassium and mixtures thereof, whereby the cesium is taken up by the modified tobermorite; separating the tobermorite containing the cesium from the liquid; mixing the tobermorite containing the cesium with portland cement to form a mixture containing up to about 40 weight percent tobermorite; and solidifying the mixture, thereby fixing the cesium ions for long-term storage. 2. The method of claim 1 wherein the tobermorite contains from about 1 to about 15 weight percent aluminum. 3. The method of claim 2 wherein the waste solution may contain cesium, lead, rubidium, cobalt and cadmium ions. 4. A method of fixing radioactive cesium for long-term storage comprising: 5. The method of claim 4 wherein the tobermorite contains from about 1 to about 15 weight percent aluminum.
048266515	claims	1. Method for assisting the loading of a reactor core with new and/or irradiated elongated fuel assemblies, which comprises inserting a fuel assembly with a fuel assembly carrier from a storage pit into a grid position of a reactor core with a refueling machine having propelling equipment, supplying an actual position of the fuel assembly carrier to the propelling equipment of the refueling machine during movement of the refueling machine between the storage pit and the reactor vessel, comparing the actual position of the fuel assembly carrier with a desired position of the fuel assembly carrier to find a deviation, and carrying out a correction of the deviation in the travelling movement of the refueling machine based upon the comparison. 2. Method according to claim 1, which comprises performing the inserting step under a neutron-shielding liquid covering. 3. Method according to claim 1, which comprises transmitting a picture from a stationary television camera to a monitor, for recognizing the deviation from the desired position and ascertaining the actual position of the fuel assembly carrier. 4. Apparatus for assisting the loading of new and/or irradiated elongated fuel assemblies having fuel assembly carriers into grid positions of a reactor core in a reactor vessel of a reactor having a storage pit, a flooding pit and a lead through therebetween, comprising a refueling machine having propelling equipment for transferring a fuel assembly from the storage pit to the flooding pit and inserting the fuel assembly into a grid position, means for recognizing the actual position of the fuel assembly carrier at a measuring point in the vicinity of the lead through and for supplying the actual position of the fuel assembly carrier to said propelling equipment of said refueling machine during movement of said refueling machine between the storage pit and the reactor vessel, means for comparing the actual position of the fuel assembly carrier with a desired position of the fuel assembly carrier to find a deviation, and means for carrying out a correction of the deviation in the travelling movement of said refueling machine based upon the comparison. 5. Apparatus according to claim 3, wherein said lead through includes guide rails for a floodgate and means supported in said guide rails for holding said recognizing means at said measuring point.
058928073	summary	FIELD OF THE INVENTION The present invention relates to a nuclear fuel assembly for a pressurized water reactor and in particular to a nuclear fuel assembly having a nuclear fuel rod with a cladding tube made of a zirconium based alloy. BACKGROUND OF THE INVENTION Cladding for use in nuclear fuel rods for light water reactors functions to prevent fission products from being released from the fuel into the coolant/moderator and to prevent contact and chemical reactions between the fuel and the coolant/moderator. The cladding is required to have excellent mechanical properties and high corrosion resistance in the environment and for the conditions expected during reactor operations. Cladding is therefore required to have adequate corrosion resistance for the lifetime of the fuel rod for operation in water and steam at temperatures up to approximately 345.degree. C., adequate strength and creep behavior over the lifetime of the fuel rod, and typically have low parasitic neutron absorption for economic use of the fissionable fuel material. Common cladding materials include zirconium, zirconium alloys, and stainless steel. Zirconium based alloys in which the major component is zirconium have been used in the cladding of nuclear fuel rods or elements for several decades. Two of the most commonly used zirconium alloys that have given satisfactory performance are Zircaloy 2 and Zircaloy 4 and are described in American Society for Testing and Materials standard B350-93(1993), Standard Specification For Zirconium and Zirconium Alloy Ingots For Nuclear Application, compositions R60802 and R60804, respectively. Zircaloy 2 (composition R60802) is composed of from 1.20 to 1.70 weight percent tin, 0.07 to 0.20 weight percent iron, 0.05 to 0.15 weight percent chromium, 0.03 to 0.08 weight percent nickel, where the iron plus chromium plus nickel content is from 0.18 to 0.38 weight percent, and the balance is zirconium plus impurities. Zircaloy 4 (composition R60804) is composed of from 1.20 to 1.70 weight percent tin, 0.18 to 0.24 weight percent iron, 0.07 to 0.13 weight percent chromium, where the iron plus chromium content is 0.28 to 0.37 weight percent, and the balance is zirconium plus impurities. The maximum impurities for Zircaloy 2 and Zircaloy 4 are given in the following table which is from Table 1 of the ASTM B350-93 Standard. TABLE I ______________________________________ MAXIMUM IMPURITIES, WEIGHT % R 60802 R 60804 ______________________________________ Aluminum 0.0075 0.0075 Boron 0.00005 0.00005 Cadmium 0.00005 0.00005 Carbon 0.027 0.027 Cobalt 0.0020 0.0020 Copper 0.0050 0.0050 Hafnium 0.010 0.010 Hydrogen 0.0025 0.0025 Oxygen * * Magnesium 0.0020 0.0020 Manganese 0.0050 0.0050 Molybdenum 0.0050 0.0050 Nickel -- 0.0070 Niobium 0.010 0.010 Nitrogen 0.0065 0.0065 Silicon 0.012 0.0120 Tin -- -- Titanium 0.0050 0.0050 Tungsten 0.010 0.010 Uranium (Total) 0.00035 0.00035 ______________________________________ * When so specified in a purchase order, oxygen shall be determined and reported. Maximum or minimum permissible values, or both, shall be as specified. Although these and other alloys have provided generally adequate performance, they possess some deficiencies that have prompted further analysis and research to find alternative materials for and alternative constructions of nuclear fuel rod cladding to single walled cladding comprised of a single metal or alloy (sometimes referred to as "through" wall cladding) which does not possess both optimum strength and resistance to corrosion. Alternative constructions to single or through wall cladding for use as nuclear fuel rod cladding includes two layer or multiple layer tubing. These types of cladding have (a) an outer layer of a highly corrosion resistant alloy and (b) an inner layer that provides the bulk of the mechanical strength of the cladding. Cladding of this type, sometimes referred to as duplex cladding, with an extra low tin Zircaloy-type outer layer (nominally 0.8 wt. % tin) and a Zircaloy-4 inner layer is currently in use for nuclear fuel rod cladding. Zircaloy-4 inner layer cladding with a thin outer layer (3 to 5 mil) of various other corrosion resistant alloys has been produced and tested in-reactor. An outer layer alloy containing 0.5 wt. % tin, 0.5 wt. % iron, balance zirconium, and another outer layer alloy containing 0.5 wt. % tin, 0.5 wt. % iron, 0.2 wt. % chromium, balance zirconium have each shown exceptional corrosion performance in a high temperature pressurized water reactor. Examples of multiple layered tubing constructions and alloys for nuclear fuel rods are discussed in U.S. Pat. Nos. 5,493,592; 4,963,316; 4,735,768, which are each hereby incorporated by reference. With the higher burnups and longer in-reactor residence times that are being pursued and which, for largely economic reasons, continue to be increased, performance limits of commonly used alloys for nuclear fuel rod cladding are being reached. The corrosion resistance of the Zircaloys has been a major concern, especially in modern high coolant temperature pressurized water reactors that employ low leakage core loadings where the corrosion film on Zircaloy can build up to unacceptable levels for burnups around 50 to 60 MWd/kgU. In efforts to optimize the corrosion performance of the Zircaloys, through a reduction in the tin level, the strength and creep properties of the cladding material have thereby been diminished. For example, over the last decade the tin level of the Zircaloys used as cladding materials in nuclear fuel rods which was nominally held at approximately 1.55 wt. % has been lowered to a nominal level of approximately 1.30 wt. %. This reduction in the level of tin has resulted in substantially better corrosion performance specifically at higher burnups, but the reduction in tin has negatively impacted the mechanical properties of the cladding. Tin is a solute solution strengthening alloy element in Zircaloy and improves the strength and creep resistance of the alloy. However, lowering the tin level in Zircaloy reduces the resistance of the cladding to creepdown as well as the strength of the cladding. In attempts to overcome the limitations in the higher burnup performance of the zirconium alloys and the Zircaloys, alloy development programs have been initiated and research and development continue to this date for zirconium alloys for use as a nuclear fuel rod cladding that would have a more favorable combination of corrosion resistance, high strength and creep resistance as well as a low neutron cross section. An object of the present invention is to improve upon the nuclear fuel rod claddings produced to date by using (I) an alloy for the outer layer of a multiple layered cladding tube with exceptional in-reactor corrosion characteristics and in accordance with the present invention to utilize (II) a new alloy for the inner part of the cladding that is of substantially higher strength than Zircaloy-2 or Zircaloy-4, while maintaining low parasitic neutron absorption characteristics of the latter alloys. By using such a higher strength alloy for an inner layer of a multiple layer cladding tube, the overall cladding tube wall thickness can be reduced while still meeting the mechanical design and performance criteria of the fuel rod. By being able to reduce the cladding wall thickness, the cladding weight per unit length of cladding can be reduced and the cost of a cladding tube of a given length is reduced since less material is needed for the production of the cladding. Furthermore, by being able to reduce the cladding wall thickness, improvements in fuel cycle costs resulting from a reduction in the parasitic thermal neutron absorption can be obtained since parasitic neutron absorption for cladding of a given composition is directly proportional to the cladding wall thickness. Alloying elements with a smaller thermal neutron cross section than currently employed tin or niobium additions can reduce the parasitic neutron absorption of the alloy even further and gain additional improvements in fuel cycle costs. By using such a higher strength alloy for an inner layer of the multiple layer cladding, significant energy production cost savings can also be obtained by reducing cladding wall thickness and increasing fuel rod fissionable material weight which is achieved by being able to use larger diameter fuel pellets while maintaining a constant fuel rod outer diameter. For a given fuel rod design, the outer diameter of the cladding is primarily determined by thermal hydraulic considerations and therefore cannot readily be changed. Thin wall cladding can accommodate larger diameter fuel pellets than a thicker wall cladding of the same outside diameter. A larger diameter fuel pellet can have a lower uranium enrichment than a smaller diameter pellet to produce the same amount of energy. For slightly enriched uranium dioxide nuclear fuel, the lifetime energy production of a unit length of fuel rod is proportional to the total number of U.sup.235 atoms per unit length. Thus, for example, by using cladding with a 0.005 inch thinner wall than a thick wall design fuel rod containing 0.300 inch diameter pellets enriched to 4.00 wt. % U.sup.235, fuel pellets of 0.310 inch diameter may be used. The reduced U.sup.235 enrichment of these pellets would be ##EQU1## (where L is a unit length of fuel) to maintain approximately the same number of U.sup.235 atoms per unit length of fuel. Alternatively, by maintaining the same U.sup.235 enrichment and increasing the pellet diameter, the number of U.sup.235 atoms per unit length of fuel rod is increased and the lifetime energy production of a unit length of fuel would be increased as well. Either alternative would lead to reactor fuel cycle cost reductions by using relatively higher cost, but thin wall, multiple layer cladding compared to using thicker through wall Zircaloy cladding. SUMMARY OF THE INVENTION The present invention relates to a nuclear fuel assembly for a pressurized water reactor having a lower tie plate, a guide tube having an upper end and a lower end, the lower end connected to the lower tie plate, spacer grids spaced along the guide tube, an upper tie plate which is attached to the upper end of the guide tube, a plurality of nuclear fuel rods which are spaced radially and supported along the guide tube by the spacer grids, at least one of the plurality of nuclear fuel rods having a metallic tubular fuel rod cladding containing nuclear fuel pellets therein and end sealing means thereon to hermetically seal the nuclear fuel pellets within the metallic tubular fuel rod cladding, wherein at least one of the nuclear fuel rods is comprised of an elongated hollow metallic tubular cladding for containing a nuclear fuel, the tubular cladding comprising an outer tubular layer having an outer wall and an inner wall, the outer tubular layer formed of a zirconium alloy, an inner tubular layer bonded to the inner wall of the outer tubular layer and formed of a high strength zirconium alloy consisting essentially of molybdenum and 3 to 6 weight percent bismuth, the balance zirconium, a body of nuclear fuel material disposed in the tubular cladding, and sealing means at both ends of the tubular cladding for hermetically sealing the metallic tubular cladding. In another preferred embodiment, a nuclear fuel assembly for a pressurized water reactor is provided comprising a lower tie plate, a guide tube having an upper end and a lower end, the lower end connected to the lower tie plate, spacer grids spaced along the guide tube, an upper tie plate which is attached to the upper end of the guide tube, a plurality of nuclear fuel rods which are spaced radially and supported along the guide tube by the spacer grids, at least one of the plurality of nuclear fuel rods comprising a metallic tubular fuel rod cladding containing nuclear fuel pellets therein, and having end sealing means thereon to hermetically seal the nuclear fuel pellets within the metallic tubular fuel rod cladding, wherein at least one of the nuclear fuel rods is comprised of an elongated hollow metallic tubular cladding for containing a nuclear fuel, the tubular cladding comprising an outer tubular layer having an outer wall and an inner wall, the outer tubular layer formed of a zirconium alloy, an inner tubular layer bonded to the inner wall of the outer tubular layer and formed of a high strength zirconium alloy consisting essentially of 1.5 to 6 weight percent bismuth and 1 to 4 weight percent tin, the balance zirconium, a body of nuclear fuel material disposed in the tubular cladding, and sealing means at both ends of the tubular cladding for hermetically sealing the metallic tubular cladding. In another preferred embodiment, a nuclear fuel assembly for a pressurized water reactor is provided comprising a lower tie plate, a guide tube having an upper end and a lower end, the lower end connected to the lower tie plate, spacer grids spaced along the guide tube, an upper tie plate which is attached to the upper end of the guide tube, a plurality of nuclear fuel rods which are spaced radially and supported along the guide tube by the spacer grids, at least one of the plurality of nuclear fuel rods comprising a metallic tubular fuel rod cladding containing nuclear fuel pellets therein, and having end sealing means thereon to hermetically seal the nuclear fuel pellets within the metallic tubular fuel rod cladding, wherein at least one of the nuclear fuel rods is comprised of an elongated hollow metallic tubular cladding for containing a nuclear fuel, the tubular cladding comprising an outer tubular layer having an outer wall and an inner wall, the outer tubular layer formed of a zirconium alloy, an inner tubular layer bonded to the inner wall of the outer tubular layer and formed of a high strength zirconium alloy consisting essentially of 1.5 to 3 weight percent bismuth, 0.5 to 3 weight percent niobium, 0.5 to 1.5 weight percent molybdenum, the balance zirconium, wherein the sum of the niobium and molybdenum is greater than 1.5 weight percent, a body of nuclear fuel material disposed in the tubular cladding, and sealing means at both ends of the tubular cladding for hermetically sealing the metallic tubular cladding.
H00002356	claims	1. In a toroidal fusion plasma device, including inner and outer confining walls, and a particle beam source for injecting a particle beam into said toroidal fusion plasma device, an arrangement for in-situ determination of species energy yields of said injected particle beam, said arrangement comprising: (a) beam stop means having a target surface, and predetermined geometry and material properties, said beam stop means positioned on said inner confining wall such that said injected particle beam is incident on said beam stop and said incident particle beam is elastic Rutherford backscattered by said beam stop; and (b) particle energy analyzing means positioned on the outboard side of said toroidal fusion plasma device, for obtaining an energy yield characteristic response of said elastic Rutherford backscattered beam particles such that the species energy yields of said injected particle beam is determined from a Rutherford backscattering analysis. (a) elastic Rutherford backscattering said particle beam by a beam stop, positioned on said inner confining wall, having a target surface and predetermined geometry and material properties; (b) obtaining an energy yield characteristic response of said elastic Rutherford backscattered beam particles; and (c) determining the species energy yields of said injected particle beam from a Rutherford backscattering analysis of the obtained energy yield characteristic response and said predetermined geometric and material properties of said beam stop. 2. In a toroidal fusion plasma device which includes inner and outer confining walls, and a particle beam source for injecting a particle beam into said toroidal fusion plasma device, a method for in-situ determination of species energy yields of said injected beam, said method comprising: 3. The arrangement of claim 1 wherein said characteristic response consists essentially of a near-surface approximation of the characteristic response of the back-scattering of said beam particles by said beam stop. 4. The arrangement of claim 1 wherein said particle energy analyzing means further comprises means for simultaneously detecting the energies of said backscattered beam particles corresponding to a plurality of target positions on said target surface from which said particle beam is backscattered. 5. The arrangement of claim 1 wherein said particle energy analyzing means comprises a charge-exchange analyzer which electrostatically determines the energy of said backscattered beam particles. 6. The arrangement of claim 1 wherein the toroidal fusion plasma device further includes inner wall armor and wherein said beam stop comprised said inner wall armor of a fusion plasma device.
053234278	abstract	A nuclear reactor containment arrangement has a permanent seal ring across the thermal expansion gap defined by the reactor vessel wall and the containment wall. The seal ring allows for lateral translation of the reactor vessel, in addition to radial and axial expansion and contraction. The seal ring features an outer attachment arrangement characterized by a radially corrugated cylinder extending vertically between two annular Belleville plates having different diameters. The corrugations of the cylinder extend longitudinally between the cylinder edges, and flex during lateral movement of the reactor vessel relative to the containment wall. The Belleville plates accommodate axial expansion and contraction of the vessel, and an inner attachment arrangement accommodates radial expansion and contraction.
	summary
063255383	description	DETAILED DESCRIPTION OF THE INVENTION FIGS. 1, 3 and 23 show generally the preferred embodiment of the apparatus of the present invention designated generally by the numeral 10. In FIG. 1, the shield is removed for clarity. The apparatus 10 of the present invention provides a shielded x-ray apparatus that includes a frame 11 having a lower end 12 that rests upon a supporting surface such as a floor and an upper end portion 13 that supports a table 14. The table 14 has head and foot end portions 15, 16 respectively and an upper surface 17. The upper surface 17 is receptive of patient 18 as shown. A suitable attachment such as a conventional sliding rail attachment secured by bolting, riveting, welding or the like can be used to form an attachment at 19 between frame 11 and table 14. An x-ray unit 20 includes a base 21 having a vertical section 22, horizontal section 23 and superstructure 26. The superstructure 26 is attached to vertical section 22 with support beam 25. Arrow 34 indicates that beam 25 can be rotatably supported by horizontal section 23 using a motor drive, for example. Support beam 25 pivotally supports superstructure 26 at connections 24 (pinned, for example). Superstructure 26 is comprised of inclined arms 30, 31 and beams 37, 38 that can be generally horizontally positioned. Pinned connections 27, 28 can be used to join the arms 30, 31 to the beams 37, 38 of superstructure 26. Each of the beams 37, 38 supports a yoke. Beam 37 supports yoke 32. Beam 38 supports yoke 33. The beams 37, 38 are supported by inclined arms 30, 31. Pinned connections 27, 28 can be used to join each of the beams 37, 38 to the respective end portions of inclined arms 30, 31 as shown in FIGS. 1 and 3. A pair of monitors 39, 40 can be positioned next to table 11 so that a radiologist can view the monitors 39, 40. Monitors 39, 40 are supported by a frame such as monitor support 41 shown in FIGS. 1 and 3. In FIG. 1, reference line 42 indicates the path that x-rays travel when emitted by radiation generator 36 in the direction of camera 35 which contains film. A pinned connection 45 can be used to join camera 35 to upper yoke 32. Similarly, pinned connection 46 can be used to join x-ray generator 36 to lower yoke 33. Arrows 43, 44 in FIGS. 1 and 3 indicate schematically the pivotal movement of camera 35 and x-ray generator 36 respectively relative to superstructure 26. Arrow 29 in FIGS. 1 and 3 schematically indicates the adjustable movement in a fore and aft direction relative to support 25 of arms 30 and 31. In this fashion, the superstructure 26 can be used to move the position of both camera 35 and x-ray generator 36. A telescoping mechanism (not depicted), located either at arms 30, 31 or at camera 35, can be used to vary the distance between camera 35 and the patient 18. The present invention has particular utility to cardiac catheterization procedures. During cardiac catheterization and intervention, procedures for which the illustrated equipment is typically used, a physician advances catheters into the patient's heart, usually through veins and arteries cannulated in the patients groin or elbow crease. These catheters are then used to inject contrast dye into the patient's cardiac chambers or blood vessels surrounding the heart, and small steerable and/or implantable tools, such as balloons, stents, and rotablators, are used to remove or modify pathologic narrowings of coronary arteries or heart valves. While most of the radiation beam 60 (see FIG. 2) emitted by the generator 36 crosses the table 14 and the patient 18, and ultimately enters the camera 35, a sizable portion is deflected by the different media it encounters on the way, as shown schematically in FIG. 2, leading to scatter radiation, depicted by dotted lines 61. This scatter radiation 61 poses health hazards, most notably the risk of cancer, leukemia, and cataract formation in the eye. Patients, while directly exposed to the X-ray beam 60, are currently felt to be only at moderate risk because the time of exposure is limited. On the other hand, physicians and allied health personnel assisting in the catheterization laboratory are repeatedly exposed to cumulative scatter radiation in doses inverse to the distance from the source. This scatter radiation is confined by the shield 50. Shield 50 consists of a rectangular middle portion 47, located between upper shield section 52 (attached to the camera 35) and lower shield section 53 (attached to the radiation generator). Attention is first directed to the upper and lower shield sections. Each of these shield sections (52, 53) has the shape of a truncated pyramid with four lateral walls 56. Each lateral wall consists of two half pieces 58, 59, which in turn are formed by a plurality of horizontal shield segments 57, joined by hinges 62. FIGS. 5a-5d illustrate this slidable joining of half pieces 58 and 59. FIG. 5c shows a single half piece. The flexible connection of horizontal shield segments 57 through hinges 62 allows for the extension or reduction of the height of the half piece. The two opposing half pieces 58, 59 are connected slidably through projections 64 on half piece 58 which engage matching slots 63 on half piece 59. In so engaging each other, the two opposing halfpieces show a section of overlap 151. This width of this section of overlap 151 will vary, depending on the degree of engagement between the two opposing half pieces 58 and 59. In FIG. 5a, the two half pieces are deeply engaged, forming a large section of overlap 151. In FIG. 5b, the two half pieces are barely joined, forming a small section of overlap 151. This overlapping of the opposing half pieces 58 and 59 is schematically illustrated in FIG. 5d. FIGS. 6a-6b illustrate in detail the projections 64 on half pieces 58, matching the slots 63 on half pieces 59. Also seen in detail is the flexible nature of hinges 62 joining horizontal shield segments 57. It stands to reason that this construction of each of the lateral walls 56 of shield sections 52, 53 allows these shield sections to assume an infinite variety of positions. They can adapt to variations in the distance from camera 35 or radiation generator 36 to middle portion 47 by varying the degree of folding between horizontal shield segments 57 at hinges 62. They can adapt to variations in the size of the middle portion 47 by varying the degree of overlap 151 through increased or decreased engagement of half pieces 58 and 59. They can adapt to rotational movements and to angulated movements of the x-ray ensemble in relation to table 14 and middle portion 47, since the degree of folding between horizontal shield segments 57 can be different for each of the four lateral walls 56. This high degree of possible adaptation is illustrated in FIGS. 4a-4d. In FIG. 7, frame 65 is shown that supports each bellows-like tapered shield sections 52 and 53. Each section 52, 53 is supported from the inside by frame 65 that includes stabilizing rods 66, consisting of three parts connected by a telescoping mechanism (66a, 66b, 66c), as illustrated in FIGS. 7 and 8. Ball joints 67 at the smaller end of these rods 66 fit into sockets 68 located at the radiation generator 36 and the camera 35 respectively. Ball joints 67 at the larger diameter ends of the rods 66 fit into sockets 68 located in connectors 69 and 70. The connectors 69, 70 of a tapered section 52, 53 are connected to each other via shafts 79 and rods 80, whereby the rods 80 lead into tubular openings of the shafts 79 (not depicted), thus forming a telescoping connection (see FIG. 7). The telescoping mechanism of these supporting rods could be moved passively through the relative movement of table, radiation generator, and camera, or better through an internal hydraulic mechanism within the telescoping rods and shafts (not depicted), allowing for active extension or retraction of the three telescoping parts of the rods 66, synchronized with the movements of the ensemble of the X-ray apparatus, using computer guidance. FIGS. 9A and 9B show the connection of tapered sections 52 and 53 to the frame 65 at rods 66. Connector rings 110 are located at each of the two telescoping joints linking the three parts 66a-66c of rods 66. FIG. 9A shows the structure of a connector ring 110. A large ring 110a is attached to the end of rod parts 66a and 66b at the point where the telescoping rod parts meet. Two radially extended struts 110b are attached with one end to the large ring 110a at an orthogonal angle to each other, and each strut has a smaller ring 110c at its other end. The rings 110c are in turn connected to metal loops 71 which are attached to the half pieces 58, 59 of a tapered section 52 or 53 by any conventional means. It should be noted that the size of the connectors shown in FIGS. 9A and 9B, and the distance from the rods to the shields, have been exaggerated for clarity. At the end opposed to the radiation generator 36, the stabilizing rods 66 supporting the lower tapered section 53 are connected via ball joints 67 to connectors 69. These connectors form the corners of a drive unit located within the table board 14, which is depicted in detail in FIGS. 10 and 11. In FIGS. 10-11, table board 14 consists of a base part 14b, and a cover part 14a. Attached to the connectors 69 are shafts 72 containing internal female threads into which thread rods 73 are connected. These threaded rods 73 are driven by small, reversible motors 74 and 75, which rotate the thread rods. Motor 74 is firmly attached to either the table base 14b, or the table cover 14a, whereas motor 75 is not attached to either. The two ends of threaded rods 73 emanating from the motors 74 and 75 have opposing helix directions, such that rotation of the threaded rods 73 in one direction will pull the attached shafts 72 and cubic connectors 69 towards the center of the table, whereas rotation in the opposite direction will push the attached shafts 72 and cubic connectors 69 away from the center of the table. Motor 75 is also attached to an axial shaft 77, which contains an internal female threaded portion into which an axial threaded rod 78 connects. This threaded rod 78 is connected at one end to axial shaft 77, and at its other end to a third motor 76, which is attached firmly to the same portion of the table as motor 74. Rotation, by motor 76, of threaded rod 78 in one direction will pull the assembly of motor 75, threaded rods 73, shafts 72, and connectors 69 towards one end of the table, whereas rotation into the opposite direction will move these parts towards the opposite end of the table. The shafts 72 glide on conventional ball-bearings located in the two side openings 14c and 14d of the table base 14b. In addition, motors 74-76 can be connected to a source of electrical power via conventional electrical power cords 101. FIGS. 12A and 12B show in detail the rod connections of a frame 65 that supports a tapered section 52, 53 and attachment to the drive unit. FIG. 12A shows a view of motor 74, connected via threaded rods 73 and shafts 72 to the connectors 69. Shafts 79 and rods 80 connect the four connectors 69 (only two of which are shown in FIG. 12A), thus forming a rectangle, as also depicted in FIG. 7. FIG. 12B shows a single connector 69, viewed from a different angle, with attached shafts 72 and 79 and rods 80. FIG. 13A depicts the framework of the middle portion or rectangular housing 47 of shield 50. Angle struts 81 are connected via locking bolts 82 to holes 83 in the connectors 69 and 70. A conventional electric power lock mechanism allows for manual closure and central closure or release of the locking bolts 82. FIGS. 14A-14C show different views of the upper and lower ends of angle struts 81, showing locking bolts 82 both in the locked and the released position. Side struts 84 (FIG. 13A) are attached to the angle struts 81 and lead into hollow bars 85, a portion of which is shown enlarged in FIG. 13B, each of which contain two parallel tracks 86 or apertures into which the side struts 84 connect. This allows the ends of the side struts 84 of the opposing angle struts 81 to move in adjacent, parallel tracks 86. The hollow bars 85 are attached to the upper and lower rims of plates 87 which contain a large rectangular opening 92. A row of conventional female snap-fasteners 88a are located around opening 92. Hollow bars 89 are attached to the angle struts 81 in a direction orthogonal to the side struts 84. U-shaped bars 90 fit into the hollow bars 89. The connection of U-shaped bars 90 and hollow bars 89 is such that the ends of U-shaped bars 90 move freely within the hollow bars 89. A compression spring (not depicted) is located within each of the hollow bars 89, and this spring pushes U-shaped bars 90 away from angle struts 81. The length of these springs is such that U-shaped bars 90 and hollow bars 89 will not completely separate. Note that the vertical portion of each U-shaped bar 90 contains a row of conventional female snap-fasteners 88a similar to the snap fasteners 88a on elements 87. Folding, X-ray impermeable screens 91, containing small, slit-like openings 111 located on the upper and lower ends of their folding segments 88 are attached to the middle, rectangular housing portion 47. Side struts 84, the hollow bars 89, and the horizontal portions of U-shaped bars 90 thread through the openings 111, as shown in FIGS. 15A-15C, thus allowing horizontal movements of the screens. The vertical ends of the screens are attached firmly to the vertical portion of the U-shaped bars 90, the vertical rims of angle struts 81, and the vertical rims of plates 87, by any conventional means. This ensemble is depicted in FIGS. 16A and 16B, and forms the side wing of middle portion 47. FIGS. 17A-17C depict "cuffscreens", which are rectangular pieces of flexible, stretchable, radio opaque material with an opening in the middle. These cuff screens can be opened and closed by means of VELCRO hook and loop type fasteners 95a, 96a, 97a, which lead from the openings 95b, 96b, 97b to the upper rim of the cuff screen. The cuff screens fit snugly around those parts of the patient's body that are leaving the middle housing portion 47 (see FIG. 3). Needed are one collar screen 95 for the patient's neck, one belt screen 96 for the waist, and two sleeve screens 97 for the arms. The outer rim of the cuff screens has a rim of male snap-fasteners 88b, which engage the female snap fasteners 88a on elements 81, 87, 90, to allow attachment to the framework of middle housing portion 47. FIG. 18 shows the middle housing portion 47, seen from the head 15 end of the table board 14. FIGS. 19a, 19b and 20 show an armboard 98, which is hooked onto each of the two side wings of middle housing portion 47. These armboards 98 consist of a support rest 99, two rods 100 which slide into receiving tubes 109 within support rest 99 (thus allowing for adjustment of the armboards 98 to the varying distances between angle struts 81), and two side triangles 102, which attach via hooks 104 to sleeves 103 on angle struts 81 (see FIG. 13A). FIG. 20 shows a horizontal cut of the key elements of the framework of middle portion 47 attached to the patient. FIGS. 21A and 21B depict the connection between the upper tapered section 52 and the upper portion of middle housing portion 47. The lowermost segments of section 52 are attached to connectors 70. The upper ends of angle struts 81 are attached to connectors 70 by passing bolts 82 through holes 120 located in shield section 52, and then into holes 83 of connectors 70. One half of the lowermost segments of shield section 52 has a row of female snap fasteners 88a. Flexible collar screens 95 attach with male snap fasteners 88b to the female snap fasteners 88a located on the vertical portions of U-shaped bars 90 and the lowermost segments of shield section 52. Seen from the foot end of the table, the view would be similar, except that belt screen 96 would replace collar screen 95. FIG. 22 depicts the connection between the upper portion of tapered section 53 and the lower portion of middle housing portion 47. The uppermost segments of tapered section 53 are attached firmly to the lower half of connectors 69. The lower ends of angle struts 81 are connected to the upper half of connectors 69 by inserting locking bolts 82 into holes 83. FIG. 23 is a side view of the invention, showing the connection of the lower portion of shield 50 to radiation generator 36 via a special connection piece 93. Connection piece 93 is rectangular at its base, where it is firmly attached to the lowermost folding segment of tapered section 53. It is circular at its lower end, where it is firmly attached to the upper rim of radiation generator 36 at attachment, completely enclosing the connection of rods 66, ball joints 67, and joint sockets 68. A similar connection piece 94 is firmly attached with its rectangular base attachment 54 to the uppermost folding segment of tapered section 52, and with its circular end connected to the lower rim of camera 35, again covering the connection of parts 66, 67, 68. Both connection pieces 93 and 94 are made of radio impermeable material. While many different variations of this design are possible to achieve the radiation shielding goals (the depicted design is only one of many different options), the unique new feature of this invention is the near complete isolation of the radiation space between the generator 36 and the camera 35 through a mobile shield that always adapts to the varying geometric relationships between the X-ray apparatus and the different organs in the patient's chest targeted for examination. The radiation beam 60 in its various positions in relation to the patient is always enclosed, allowing no scatter radiation 61 to escape. An important point is that the material and structure of shield sections 52, 53 allow for asymmetry of the shields, which will occur as the X-ray apparatus moves, that it allow for a variation in the distance between radiation generator 36 and camera 35, and for a variation in the distance between either of these parts and table board 14. The waist collar 96 has to be quite flexible, and the entire lower portion of this part has to be covered in sterile sheets, in order to allow the operator to move the region of vascular access into the radiation field for x-ray guidance, should this become necessary due to access problems. The support structure of shield 50 has to be sufficient to allow the free passage of the radiation beam on its way from the radiation generator to the camera, without swinging of the shields into the trajectory of the X-ray beam when the shields assume angulated positions. This is achieved through the stabilizing rods 66 in this design, but other mechanisms can be used to achieve the same effect. The middle portion has been depicted as a simple assembly of a frame structure with folding screens and cuff screens with sleeve openings, but multiple variations of this design are also possible. The present invention will allow open access to the patient's head, arms and groins. The only remaining leaks for scatter radiation are those parts of the patient's body which are emanating from the isolated radiation space. A telescoping mechanism located at camera 35 could be used to allow the camera to protrude to varying degrees directly into the shielded space formed by tapered section 52. Several other features are important. The device of the present invention can be disassembled very quickly, so that effective cardiopulmonary resuscitation can be rendered without delay in the case of cardiac arrest. Sensitive radiation detection devices have to be installed outside the confinements of the radiation field isolator, in order to prevent accidental exposure of the operators to scatter radiation leaks. While the device presented is intended for diagnostic and therapeutic procedures using radiation for imaging purposes, minor modifications in the design would easily allow to use this device for radiation therapy. Although the Radiation Field Isolator and the method of using the same according to the present invention have been described in the foregoing specification with considerable details, it is to be understood that modifications may be made to the invention which do not exceed the scope of the appended claims, and modified forms of the present invention done by others skilled in the art to which the invention pertains will be considered infringements of this invention when those modified forms fall within the claimed scope of this invention. The following is a list of parts and materials suitable for use in the present invention: PARTS LIST Part Number Description 10 shielded x-ray apparatus 11 frame 12 lower end portion 13 upper end portion 14 table 14a cover part 14b base part 14c opening 14d opening 15 head end 16 foot end 17 upper surface 18 patient 19 attachment 20 x-ray unit 21 base 22 vertical section 23 horizontal section 24 pinned connection 25 support beam 26 superstructure 27 pinned connection 28 pinned connection 29 arrow 30 arm 31 arm 32 upper yoke 33 lower yoke 34 arrow 35 camera 36 radiation generator 37 beam 38 beam 39 monitor 40 monitor 41 monitor support 42 reference line 43 arrow 44 arrow 45 pinned connection 46 pinned connection 47 rectangular housing or middle portion 48 wall 49 flexible panel 50 shield 51 opening 52 upper tapered section 53 lower tapered section 54 upper attachment 55 lower attachment 56 flexible lateral wall 57 horizontal shield segments 58 half piece 59 half piece 60 radiation beam 61 scatter radiation 62 hinge 63 slotted receptacle 64 projecting member 65 frame 66 rod 66a telescoping part 66b telescoping part 66c telescoping part 67 ball joint 68 socket 69 connector 70 connector 71 loop 72 shaft 73 threaded rod 74 reversible motor 75 reversible motor 76 reversible motor 77 axial shaft 78 threaded rod 79 shaft 80 rod 81 angle strut 82 locking bolt 83 hole 84 side strut 85 bar 86 parallel tracks 87 plate 88 folding segment 88a female snap fastener 88b male snap fastener 89 bar 90 bar 91 screen 92 opening 93 connecting piece 94 connecting piece 95 collar screen 95a VELCRO fastener of collar screen 96 belt screen 96a VELCRO fastener of belt screen 96b opening of belt screen 97 sleeve screen 97a VELCRO fastener of sleeve screen 97b opening of sleeve screen 98 armboard 99 support rest 100 rod 101 power cord 102 side triangle 103 sleeves 104 hooks 109 receiving tubes 110 connector ring 110a ring 110b strut 110c small ring 111 slit like opening 120 hole 151 section of overlap The forgoing embodiments are presented by way of example only; the scope of the present invention to be limited only by the following claims.
06256594&	claims	1. A machine fault monitoring apparatus having, for each of a plurality of machines, fault detection means for detecting various faults occurring during the operation of the machines, and remote monitoring station for monitoring operating status of the plurality of machines by collecting fault detection data locally detected by the fault detection means of the plurality of machines, wherein each machine comprises: operating parameter detection means for locally sequentially detecting various types of operating parameter values which change during the operation of said machine; history data update means locally updating fault detection history data every time a fault is detected by said fault detection means during the operation of said machine; determination means for locally determining, based on said history data and the value of the fault, whether or not to transmit to said monitoring station sequential values of the operating parameters within a specified period of time around a point in time at which the fault was detected, in cases where said fault was detected by said fault detection means during the operation of said machine; transmission means for transmitting to said remote monitoring station, in cases where it has been determined locally by said determination means that a transmission should be made, a type of fault that was detected, a value detected by said operating parameter detection means at the point in time at which the fault was detected, as well as the sequential values of said operating parameters within the specified period of time around the point in time at which the fault was detected, and also for transmitting to said remote monitoring station, in cases where it has been determined by said determination means that no transmission should be made, the type of fault that was detected and the value detected by said operating parameter detection means at the point in time the fault was detected. operating parameter detection means for locally sequentially detecting various types of operating parameter values which change during the operation of said vehicle; history data update means for locally updating fault detection history data every time a fault is detected by said fault detection means during the operation of said vehicle; transmission determination means for locally determining based on said history data and the value of the fault, whether or not to transmit to said remote monitoring stations sequential values of the operating parameters within a specified period of time around a point in time at which the fault was detected in cases where said fault was detected by said fault detentions means during the operation of said vehicle; and fault data transmission means for transmitting to said remote monitoring stations, in cases where it has been determined by said transmission determination means that a transmission should be made, a type of fault that was detected, a value detected by said operating parameter detection means at the point in time at which the fault was detected, as well as the sequential values of said operating parameters within the specified period of time around the point in time at which the fault was detected, and also for transmitting to said remote monitoring station, in cases where it as been determined by said transmission determination means that no transmission should be made, the type of fault that was detected and the value detected by said operating parameter detection means at the point in time the fault was detected; and wherein said remote monitoring station comprises: requested signal transmission means for transmitting to the vehicle, in cases where no relation was determined by said relation determination means, a signal requesting that values detected by said operating parameter detection means to be transmitted to said monitoring stations for a given period of time. 2. The machine fault monitoring apparatus according to claim 1, wherein said transmission means searches for a frequency of the detected fault based on said history data, and transmits data indicating the fault frequency to said remote monitoring stations. 3. A vehicle fault monitoring apparatus having, for each of a plurality of vehicles, fault detection means for locally detecting various faults occurring during the operation of the vehicles, and a remote monitoring stations for monitoring operating status of the plurality of vehicles by collecting fault detection data detected by the fault detection means of the plurality of vehicles, wherein each vehicle comprises: 4. The vehicle fault monitoring apparatus according to claim 3, wherein said fault data transmission means searches for a frequency of the detected faults based on said history data, and transmits data indicating the fault frequency to said remote monitoring stations.
	claims	1. An apparatus for a computed tomography (CT) system, comprising:a pre-object filter configured to rotate about an axis of the pre-object filter and shape a profile of radiation attenuation in a fan-angle wherein:the pre-object filter comprises a core between a first end and a second end of the pre-object filter, wherein the core has a circumference that is smaller than a circumference of the first end and a circumference of the second end, wherein the first end, the second end, and the core define an aperture;wherein a width of the aperture at a first rotation angle of the pre-object filter about a longitudinal axis of the pre-object filter differs from a width of the aperture at a second rotation angle of the pre-object filter about the longitudinal axis of the pre-object filter, and a depth of the aperture at the first rotation angle is substantially equal to a depth of the aperture at the second rotation angle. 2. The apparatus of claim 1, wherein the core further comprises a curved sidewall wherein the curved sidewall defines the aperture. 3. The apparatus of claim 2, wherein the curved sidewall is characterized by a parabolic shape in a cross-section of the core. 4. The apparatus of claim 1, wherein a change in the width of the aperture from the first rotation angle to the second rotation angle is smooth. 5. The apparatus of claim 1, wherein a change in the width of the aperture from the first rotation angle to the second rotation angle comprises at least one incremental step. 6. The apparatus of claim 1, wherein a curved sidewall extends substantially 360 degrees around the core defining the aperture substantially 360 degrees around the core. 7. The apparatus of claim 1, wherein the aperture has a third width at a third rotation angle of the pre-object filter about the longitudinal axis of the pre-object filter, wherein the third width is substantially equal to the first width and the third rotation angle is substantially 180 degrees from the first rotation angle about the longitudinal axis of the pre-object filter. 8. The apparatus of claim 1, wherein the width at the first rotation angle is a narrowest portion of the aperture and the width at the second rotation angle is a widest portion of the aperture. 9. The apparatus of claim 8, wherein the first rotation angle is 0 degrees and the second rotation angle is 90 degrees about the longitudinal axis of the pre-object filter. 10. The apparatus of claim 1, further comprising a passage defined within a center of the pre-object filter extending from the first end to the second end along the longitudinal axis of the pre-object filter. 11. A computed tomography (CT) system, comprising:a radiation source;a detector array;a rotating gantry configured to rotate the radiation source and the detector array about an object under examination; anda pre-object filter positioned between the radiation source and the object and configured to shape a profile of radiation attenuation as a function of a profile of the object, wherein:the pre-object filter comprises a core extending between a first side and a second side,a channel is defined in the core extending in a direction substantially parallel to an axis of rotation of the rotating gantry, andthe channel has a first width in a first position and a second width in a second position, wherein the second width is larger than the first width. 12. The computed tomography system of claim 11, wherein a cross-sectional shape of the channel is characterized by a curve. 13. The computed tomography system of claim 12, wherein the curve is substantially parabolic. 14. The computed tomography system of claim 11, wherein the core is substantially planar. 15. The computed tomography system of claim 11, wherein the core is substantially cylindrical. 16. The computed tomography system of claim 15, wherein the first position is a first angle of rotation of the pre-object filter about a longitudinal axis of the pre-object filter and the second position is a second angle of rotation of the pre-object filter about the longitudinal axis of the pre-object filter. 17. The computed tomography system of claim 16 wherein the first angle of rotation is about 0 degrees about the longitudinal axis of the pre-object filter and the second angle of rotation is about 90 degrees about the longitudinal axis of the pre-object filter. 18. A method for imaging a patient, comprising:acquiring a profile of the patient, the profile describing one or more features of the patient;performing an imaging scan on the patient; andshaping, as a function of the profile of the patient, a profile of radiation attenuation in a fan-angle direction to affect an amount of radiation attenuated in the fan-angle direction, wherein the shaping comprises:rotating a pre-object filter about a filter axis, wherein the pre-object filter is configured to shape the profile of radiation attenuation, wherein:the pre-object filter comprises a core extending between a first side and a second side,a channel is defined in the core extending in a direction substantially parallel to an axis of rotation of a rotating gantry, andthe channel has a first width in a first position and a second width in a second position, wherein the second width is larger than the first width. 19. The method of claim 18, wherein the filter axis is substantially parallel to the fan-angle direction.
041949480	claims	1. A locking device for supporting and locking a nuclear fuel assembly in place upon a fuel core support plate, the fuel assembly having an elongated, generally tubular configuration and the support plate having a vertically arranged bore for receiving the fuel assembly, comprising a suppport and locking sleeve having a continuous tubular portion disposed between the fuel assembly and the support plate bore, an upper portion of said sleeve having a plurality of upwardly extending fingers each forming at its upper end a contact surface facing radially inwardly and upwardly and a radially outwardly facing surface, a lower portion of said sleeve having a plurality of axially extending fingers each forming at its lower end a contact surface facing radially outwardly and downwardly and an opposed radially inwardly facing surface, means on the fuel assembly forming an annularly tapered surface arranged for engagement with said contact surfaces on said upwardly fingers, and means on the support plate bore forming an annularly tapered surface arranged for engagement with said contact surfaces on said downwardly extending fingers, said upper and lower fingers being sized so that their opposed surfaces bear on the support plate bore and the fuel assembly surface respectively while their contact surfaces are in engagement with the tapered surfaces on the fuel assembly and support plate bore respectively. 2. A locking device according to claim 1 further comprising means for restricting undesired upward movement of the fuel assembly. 3. A locking device according to claim 2 wherein said restricting means may be disengaged to permit initial upward movement of the fuel assembly for facilitating installation and removal of the fuel assembly from the bottom of the support plate. 4. A locking device according to claim 1 wherein the fuel assembly is of a type adapted for use in a gas cooled nuclear reactor and comprising means for restricting the bypass flow of coolant gas through the support plate bore around the fuel assembly. 5. A locking device according to claim 1 wherein a flow regulating means is adjustable to vary the amount of coolant gas flow permitted through the fuel assembly. 6. A locking device according to claim 1 wherein said lower fingers are pre-stressed so that they tend to move radially outwardly against the support plate bore in order to facilitate installation and removal of the fuel assembly from the bottom of the support plate. 7. A locking device according to claim 6 wherein the upper fingers are pre-stressed so that they tend to act radially inwardly against the fuel assembly in order to prevent accidental dropping of the fuel assembly from the support plate, the fuel assembly and said sleeve being movable upwardly relative to the support plate bore so that said upper fingers may be retracted outwardly in order to permit installation and removal of the fuel assembly. 8. A locking device according to claim 1 wherein said continuous tubular portion of said sleeve forms a passage for communicating vented gases from an outlet of the fuel assembly and further comprising seal rings arranged above and below said passage for limiting the flow of said vented gases. 9. A locking device according to claim 1 and further comprising a combination of key means and key way means formed at the lower ends of the support plate bore and fuel assembly to control angular alignment of the fuel assembly within the support plate bore.
054141974	claims	1. A method of containing and isolating toxic or hazardous wastes comprising the steps of: removing liquid phase from a toxic or hazardous waste to form a dry waste salt; mixing said dry waste salt into a molten thermosetting polymer to form a first mixture; forming pellets of said first mixture; allowing said molten thermosetting polymer to thermoset; heating and coating said pellets with a powdered material which is compatible with a cementitious mixture to form coated pellets; mixing said coated pellets with said cementitious mixture to form a polymer-aggregate concrete; and allowing said polymer-aggregate concrete to harden. a matrix comprising a cementitious mixture surrounding a plurality of waste pellets therein, each said pellet comprising a salt of a toxic or hazardous waste encapsulated in a thermosetting polymer and coated with a powdered or granular material which is compatible with said cementitious mixture. 2. A method as claimed in claim 1, wherein said pellets have a density and said cementitious mixture has a density which is greater than the density of said pellets, said method further including the step of disposing said polymer-aggregate concrete in a mold and spinning said mold to centrifugally force said cementitious mixture toward an outside portion of said mold and allowing said polymer-aggregate concrete to harden. 3. A method as claimed in claim 1, wherein said thermosetting polymer comprises at least one member selected from the group consisting of asphalt, polyethylene and elemental sulfur. 4. A method as claimed in claim 1, wherein said cementitious mixture comprises portland cement. 5. A method as claimed in claim 1, wherein said cementitious mixture is a mortar containing portland cement and sand. 6. A method as claimed in claim 1, wherein said powdered material comprises at least one member selected from the group consisting of ground slag, sand, fly ash, fumed silica, calcium carbonate, portland cement, ground limestone, ground clay. 7. A method as claimed in claim 1, wherein said powdered material comprises portland cement. 8. A method as claimed in claim 1, wherein said powdered material comprises ground clay. 9. A method as claimed in claim 1, wherein said powdered material comprises calcium carbonate. 10. A method as claimed in claim 1, wherein said thermosetting polymer is an asphalt and said powdered material is portland cement. 11. A method as claimed in claim 1, wherein said pellets are substantially spherical and have an average diameter of between about 0.5 inch and about 2.0 inches. 12. A wasteform produced by the method according to claim 1. 13. A wasteform produced by the method according to claim 2. 14. A wasteform produced by the method according to claim 3. 15. A wasteform produced by the method according to claim 6. 16. A wasteform containing and isolating toxic or hazardous wastes, said wasteform comprising: 17. A wasteform as claimed in claim 16, wherein the cementitious mixture forms a continuous coating on the outer portion of said wasteform. 18. A wasteform as claimed in claim 16, wherein said toxic or hazardous waste comprises a radioactive material.
056407025	description	DETAILED DESCRIPTION Reference is made to FIG. 1 for schematically illustrating a preferred embodiment of an improved waste disposal system 10 made according to the principles of the invention (the present application for which is a continuation-in-part of U.S. patent application Ser. No. 07/852,543, pending, incorporated by reference). In the illustrated waste disposal system 10, the volume of mixtures of solid, liquid and gaseous contaminants is diminished significantly through reduction or entertainment with a molten reducing metal, such as aluminum. In the specification and claims, the term "contaminant" is inclusive of a variety of wastes which are considered hazardous or radioactive, or both. It also includes inherently non-radioactive and non-hazardous wastes which, however, become hazardous or radioactive because of being admixed with hazardous or radioactive waste materials. For example, the solids which can be treated include non-radioactive, non-hazardous implements and materials of metal, plastic, glass, paper, or biological materials which have been used, treated with, or contaminated by radioactive materials; and with otherwise hazardous materials in the form of solvent, chemical reagents, poisons, or diseased biological materials. The hazardous materials can include a group comprising solvents, chemical reagents, poisons, or diseased biological materials and which have been used in conjunction with, or contaminated by, radioactive materials. The non-radioactive, non-hazardous waste materials may include syringes, needles, animal cages, specimen containers, glass tubes, vials, caps, tissues, towels, clothing, surgical implements, mechanical contrivances, and any other implement or device used in experimentation, industrial use or power generation using radioactive materials, and which have been or may have been contaminated by radioactive materials. The radioactive materials may be any radio-nuclide occurring naturally; those that have been produced by nuclear fission or fusion, or by particle accelerators or other artificial means. The radioactive materials may include deliberate or accidental inclusion in a solvent, reagent or biological material. For instance, the radioactivity can be attributable to nuclear fuel, uranium, or other fuel processing and radiopharmaceuticals. The hazardous solvents, chemical reagents or poisons may include: a) halogenated hydrocarbons, including polychlorinated biphenyls, chlorinated dioxins, chlorinated furans, and all aromatic and aliphatic organic compounds, solvents, insecticides or herbicides which are partially or completely chlorinated; b) hazardous halogenated or non-halogenated organic compounds containing as substituents, oxygen, nitrogen, sulfur or phosphorus, either singly or in combination with other elements; to include aldehydes, ketones, alcohols, carboxylic acids, esters, ethers, nitriles, amines, sulfides, thiols, thioketones, thiocarbonyls, mercaptans, phosphates, phosphites, phosphonates, phosphines and phosphine oxides, nitro compounds, nitroso compounds, amides, and amino acids, amino alcohols, sulfonic acids, sulfonates, and sulfones, thioamines, amino-thiols, and any other combinations of these with each other, or with other elements; c) nerve gases and other cholinesterase inhibitors; mustard gases; and other military chemical agents; d) heavy metal oxides, sulfides and selenides; e) anionic groups containing heavy metal and oxygen, sulfur or selenium; f) phosphorus and selenium sulfides and oxides; g) oxidizing anionic groups containing halogen; h) anionic groups containing sulfur or nitrogen; i) hazardous halides; and j) cyanides. The biological materials may include tissues from animals, biological fluids, infectious bacteria, viruses, spores, or carcinogenic agents. Reference is now made back to the drawing wherein the illustrated waste disposal system 10 includes a liquid storage assembly 12. The assembly 12 receives, stores, and dispenses contaminated liquid waste materials 14. As will be discussed, a wide variety of liquid wastes are contemplated. The storage assembly 12 includes a liquid storing tank or vessel 16 of a type suitable for receiving, holding and dispensing the particular kind of liquid contaminants 14 which are to be processed by the system 10. For example, since a contaminated liquid waste material 14 may include radioactive components then the storing tank 16 would be provided with suitable shielding for such radioactivity. The storage tank 16 is associated with an appropriate pumping and valving mechanism 18 and controls (not shown) therefor which are operable for allowing controlled discharge of preselected amounts of contaminated liquid waste materials 14 to an exit pipe 22 nd then to a liquid metal chemical reactor unit 38 to be described. The storage tank 16 is provided with doors or chutes (not shown) for permitting feeding of the wastes thereinto in either batch or continuous modes. Moreover, it is suitably lined with an appropriate lining material (not shown) to handle the contaminated wastes to be processed. For instance, the liquid wastes 14 could be transformer oils contaminated by PCBs. The system 10 also includes a gas storage cylinder 13 of a type suitable for receiving, holding, and dispensing the particular kinds of gaseous contaminants 15 which are to be processed by the system 10. Suitable valving 17 allows for both liquids and gases to utilize the same injection port. The system 10 also includes a sealed hopper 26 which is used to receive, store and dispense contaminated solid waste materials 28. The sealed hopper 26 is also provided with doors or chutes (not shown) for permitting batch or continuous feeding thereinto. It is desirable to minimize the quantity of moisture present in the feedstock. The contaminated solid waste materials 28 can comprise a variety of wastes as described herein. The sealed hopper 26 is provided with an outlet that is under the control of a suitable conveying mechanism 32 which transports the same to the liquid metal chemical reactor unit 38. The conveying mechanism 32 in this embodiment is a screw-type feed conveyor which leads to the reactor unit 38. Other types of solid feeders are, of course, envisioned. The present invention may include either the liquid, gaseous or solid feed operation. A pressurized gas system 34 is provided which directs an inert gas under positive pressure to the cylinder 13, the tank 16, or to the hopper 26 so as to enhance the feeding operation and prevent back-up and primarily to purge the system when shut down for maintenance. The present invention contemplates treating in a new and improved manner not only the hazardous liquid and solid wastes that are described in U.S. Pat. Nos. 4,469,661; 4,552,667; 4,599,141; 4,666,696; and 4,695,447, but those contaminated by radioactivity from a variety of sources. A description of those materials and the reactions which occur when using heated powdered aluminum, molten aluminum and/or other like reducing metals are incorporated herein by reference. As will be explained in the following examples, the liquid aluminum A chemically reduces the non-radioactive hazardous and non-hazardous waste materials such that there results a much lower volume of wastes, contaminated only with radioactivity; thus removing the hazardous materials, and encapsulating most of the radioactive materials in final form for disposal. Referring back to the liquid metal chemical reactor, it is seen to include an appropriately environmentally sealed housing assembly 36 which defines a liquid metal application and reaction chamber 38. The application and reaction chamber 38 can be lined with a refractory ceramic material. The application and reaction chamber 38 allows introduction of the contaminated solid, gaseous and liquid wastes, as well as permits the liquid or molten metal to be applied. A furnace for heating the system 70, also supports the chamber 38, and heating may for instance be by induction coils. In the application and reaction chamber 38, molten or liquid metal A is dispensed so as to contact the external surfaces of the solid waste materials 28 as well as the vapors formed by the evaporation of the liquid waste materials 14, or the gaseous materials 13. Aluminum is preferred as a reducing metal because of its low melting point, ready availability, stability at ordinary temperatures and volatility of its anhydrous chloride salt. Other metals have some of these desirable chemical properties including alkali metals, alkaline earth metals, iron, zinc and the rare earth metals, but aluminum is more active than some and much easier and safer than others to handle and ship. The housing assembly 36 including the liquid metal application chamber 38 are suitably insulated, closed and sealed to prevent undesired escape of the contaminants and other materials as well as prevent the inclusion of oxygen. They are also suitably thermally insulated to insure that the desired chemical reactions are carried out and that there is an adequate flow of the reacted mass of metal. In addition, the application chamber 38 is radioactively shielded with a lead partition 62. For effecting efficient contact, molten metal A is pumped from a suitably heated and insulated reservoir 40 by a pumping mechanism 42 also located in the reservoir. The level of liquid aluminum A maintained in a molten bath or pool 44 is such as to prevent the escape of gaseous reaction products or the incursion of oxygen; as well as to provide adequate metal to the pump intake. The gaseous, liquid and/or solid waste materials 13, 14 and 28 entering the application and reaction chamber 38 are contacted by a shower 46 of molten metal A. The molten metal shower 46 is achieved by reason of pumped liquid metal A descending through a plurality of spaced openings 48 which are defined by an arrangement of perforated plates 50 each having a multiplicity of fine openings 48. The openings 48 are of such a size that they permit an adequate shower of liquid metal A for the purposes intended. For example, the perforated openings 48 may have sizes which range from 1/8 to 3/8 inches. It has been found that such a size range is adequate for purposes of generally evenly distributing the molten metal A to provide a circuitous path through the streams of reducing metal, to provide means for all gaseous material to react. The foregoing arrangement also allows for continuous renewal of molten metal A for effecting the desired contact. Of course, the foregoing metal application technique is but one preferred embodiment of many which may achieve the desired end of generally evenly contacting the gaseous, liquid, and solid waste materials 14, 28 with the molten metal A. For instance, the present invention contemplates effecting such contact through the utilization of a liquid curtain or flowing surfaces among other techniques. Some reacted solid contaminants, together with reacted salts of aluminum and unreacted oxides of metals more active than aluminum (Na, Ca, Mg, K, etc.) and unreacted liquid metal are carried in a stream to a return channel 54. The channel 54 is defined by a side wall of the reservoir 44 and an extended wall of the housing 36. This channel 54 generally directs the reacted solids coming from the chamber into reservoir 44 for easier subsequent removal, such as by fluxing and skimming the liquid from this channel, by means of remotely controlled apparatus (not shown). It also provides a seal to prevent the escape of gases or the incursion of oxygen. The slag and dross formed in the application and reaction chamber 38 can be skimmed from the bath 44 as necessary by a remotely operated skimmer (not shown) and placed in ingot molds (not shown) for subsequent disposal. A suitable screen or grid or the like (not shown) can be arranged in the bottom reservoir 44 near the channel 54 so as to restrain unmelted solids from blocking the channel 55 into the pump well and interfering with the pumping operation. The present invention contemplates utilization of an ingot-forming mold (not shown) into which the withdrawn slag and reacted mass can be transferred. A standard molten metal tapping mechanism generally designated by reference numeral 56 is used for removing a portion of the unreacted liquid metal and waste material reaction products from the channel 71 for forming an ingot. In this manner the reacted gaseous, solid and liquid waste materials are reacted to a less hazardous and/or innocuous state and encased in the reducing metal, thereby producing a readily disposable ingot which contains the radioactive elements from the waste materials. Those ingots formed when a radioactive melt is tapped are, of course, radioactive, so the operation must be remotely operated. This mode has the further advantage that the metallic and metal oxide residues can be cast as ingots or bricks of a size that can allow the radioactive heat to dissipate thermally. It will be appreciated that the temperature of the reservoir 40 as well as the application and reaction chamber 38 are such as to insure that the metal remains molten. For example, if the molten bath 44 is to be substantially completely aluminum, temperatures ranging from about 600.degree. C. to about 3000.degree. C. are ordinarily useful, while temperatures ranging from 780.degree. C. to about 1000.degree. C. are preferred. An eutectic melt containing 10% aluminum and 90% zinc can be used to operate at a much lower temperature. Continuous addition of aluminum would be necessary, since it is the more active of the two metals and will react preferentially. Those waste materials, especially radioactive waste, having a melting point above the temperature of the molten metal may not become alloyed with the molten metal, but rather entrained therewith. Accordingly, the unalloyed waste material will be encased in metal ingots when finally tapped and molded. As a consequence, the waste materials will be alloyed with the metal or will remain in suspension in the molten metal. Of course, by controlling the temperature of the molten metal between the melting and boiling points of the radioactive elements, such action will control whether the radioactive elements vaporize, or react with the metal and become soluble therein. Thus, it will be appreciated that in terms of the radioactive waste materials generated for example, in medical therapy, the final form of the disposable product will vary dependent on the type of radioactive material treated. The separation of such radioactive materials as iodine, gallium, cesium, strontium 90, thallium, etc., will depend upon their reactivity with molten aluminum, temperature of the molten aluminum compared to the melting temperatures of the radioactive materials and the vaporization temperatures of the radioactive elements. Any oxides of active metals will result in a dross material that is subsequently drawn-off along with aluminum oxide. It will be further appreciated that the radioactivity will reside in a few elements that will react with the aluminum accordingly. It has been found, however, that higher temperatures diminish the viscosity of the molten metal. There is greater contact between the liquid aluminum (or other metals or alloys, including for instance scrap metals) and the solid wastes being treated as well as higher rates of reduction. The molten shower 46 also presents continuously renewed reactive surfaces for contacting. For instance, the reservoir 40 need not have the configuration shown, however, the chamber 44 should be in fluid communication with the reactor unit. While this embodiment illustrates that the gaseous, liquid and solid waste materials are introduced into a reaction zone of the chamber 38 through inlets 22a an 32a, it will be appreciated that other feeding approaches are contemplated so as to insure intimate contact of the molten metal with the gaseous, solid and liquid waste materials. Referring back to the application and reaction chamber 38, it is also noted that gases will be formed as a reaction product of some of the contaminated waste materials and the liquid metal A. In these situations, reacted waste gases and particulates are vented through an appropriate vent opening 58 by means of a vent pipe 60 to a contaminant gas trap and scrubber (not shown). Such gases will comprise primarily hydrogen from the decomposition (reduction) of water and organic compounds and some hydrocarbons from the reduction of organic compounds. It is advisable to minimize moisture quantity in the feedstock to avoid excessive use of molten aluminum. In the embodiment illustrated in FIG. 1, the system includes the arrangement wherein it is mounted on a heating means 70 which may be an induction furnace or other furnace of a known type. In this arrangement the reservoir for the molten reduction metal A is connected by suitably enclosed, sealed and insulated channels 71. The furnace portion 70 will support the housing assembly 36 of the reactor unit 38. The return channel 54 formed in the housing assembly 36 communicates with channel 71 in the furnace portion 70 so as to allow circulation of the molten metal to the heating means and back to the metal pump 42. The pump 42 forces the molten metal A to and through openings 48 in the perforated plate 50 supported by baffles B. It will be noted that the skimming system (not shown) described above can be connected to the reservoir portion 44 to perform the functions described earlier. It is to be understood that, because the embodiment is designed to reduce the material of "mixed" wastes, (i.e. wastes containing both hazardous and radioactive wastes) the physical orientation of all the parts must be such as to allow for shielding for all radioactive materials in order to minimize personnel exposure to radiation. Such a configuration is shown in FIG. 1, showing radiative shield partition 62, which allows personnel to work in the vicinity of the system without undue exposure to radiation. To better understand why aluminum is preferred in the inventive system and process it will be realized that aluminum is an active reducing agent, both in aqueous systems, and in the molten state. For instance, it is capable of: a) reducing halogenated organic compounds to carbon, hydrogen or low-molecular weight hydrocarbons; forming aluminum chloride; b) reducing ethers, esters, carboxylic acids, alcohols, and carbohydrates to carbon, hydrogen, and hydrocarbons, forming aluminum oxide; c) reducing amines, ammonia, and ammonium compounds to nitrogen, hydrogen, carbon, and forming aluminum nitride; d) reducing the halogen salts (chloride, bromide, iodide, astatide) of nearly all the metals, forming aluminum halides, some of which are volatile; e) reducing sulfate, nitrate, phosphate, arsenate, selenate and the oxyacid salts of transition metals (chromate, permanganate, etc.) to form aluminum sulfide, aluminum nitride, nitrogen, phosphorus arsenic, aluminum selenide, and the elemental form of the metals, respectively; f) reducing the oxides of many metals to the metallic form, which will alloy with the aluminum (i.e. dissolve in it), and g) reducing oxy-acid and organic acid salts of most "heavy" metals, leaving either the elemental metal dissolved in the aluminum or an oxide in the dross. Moreover, the following describes certain elements whether radioactive or not, and their reaction in the presence of molten aluminum: Group IA Metals: Aluminum will react with the halides to form aluminum chloride thus reducing these on a transient basis. The metal thus formed will react immediately with any oxygen-containing compound, reducing it and forming the metal oxide, which will remain as a slag on the molten metal surface. Those in the form of salts of oxy-acids will be decomposed by the reduction of the sulfate, nitrate, etc., and react to form the oxide. Those present as oxide will be unaffected, and will merely add to the mass of slag. Group IIA: Aluminum will reduce beryllium, magnesium and calcium to the metallic form. These will remain alloyed with the aluminum and add to the reductive mass (i.e. they will serve as reducing agents, probably dissolved in the molten aluminum). Strontium, barium and radium will form oxides and remain as part of the slag on the surface of the aluminum pool. Group III: Boron, gallium, indium and thallium will be reduced and alloy with the aluminum. Scandium, yttrium, and the rare earths, (i.e. lanthanides and actinides) are similar to aluminum as reducing agents. They will either be reduced and remain alloyed with the aluminum, or form the oxide and remain in the slag. Transition Metals will be reduced to the metal, when fed to the aluminum system, regardless of their oxidation state. These will remain alloyed with the aluminum. The presence of high quantities of these in the melt may eventually require tapping the melt to preclude having to raise the temperature too high in order to keep it molten. The coinage metals, copper, silver, gold, platinum, etc., will be reduced to the metals and remain alloyed with the aluminum. Zinc, cadmium and mercury will be reduced to the elemental state. Some of the zinc may remain alloyed with aluminum. Part of the zinc, and all of the cadmium and mercury will distill from the melt to be trapped in the trap system. It is understood that radioactive elements are chemically identical to stable (non-radioactive) elements. Therefore the reactions which take place with aluminum will take place with radioactive and non-radioactive elements alike. Hereafter follows several examples relating to the system and method of the invention. It is to be understood that these examples are illustrative, rather than limiting. Examples in the cited patents are also included by reference. EXAMPLE 1 A transformer oil was heated for 30 minutes in a sealed tube with aluminum foil at 500.degree. C. This resulted in recovery of 21.5% chloride. This indicates that, in the absence of intimate contact with the solid metal, there is an appreciable time requirement. This should be obviated by the use of molten metal. EXAMPLE 2 The melting point of aluminum is 660.degree. C. Addition of zinc metal lowers the melting point to a minimum at 382.degree. C. At this point, the zinc must be 95% of the melt, and might become a major reactant, resulting in the formation of ZnCl, which would separate in the molten state. The use of an intermediate concentration of zinc could lower the temperature to obtain the optimal conversion reaction. Since the aluminum reacts preferentially, it would be possible to feed in fresh aluminum as it is removed by the reaction. EXAMPLE 3 Aluminum forms a eutectic mixture with 13% magnesium and 8% zinc. This has a minimum melting point at about 500.degree. C. It is advisable to operate at the lowest possible temperature at which the desired reactions take place efficiently. This may allow some solvents or transformer oils to pass through the system without thermal decomposition. The preponderance of aluminum in this system makes it economically desirable compared to the high zinc eutectic. EXAMPLE 4 N-Butyl alcohol was immersed in molten aluminum. A gas was generated which corresponded to a mixture of hydrogen and 1-butene. Aluminum oxide formed. EXAMPLE 5 Dimethyl phthalate is substituted for the butanol in Example 4 with the result that the dimethyl phthalate is destroyed and hydrocarbon gas and hydrogen are produced. EXAMPLE 6 Acetonitrile was destroyed by immersion in molten aluminum. No cyanide or cyanogen was detected in either the evolved gas or in the cooled melt. EXAMPLE 7 Naphthylamide was immersed in molten aluminum and reacted with evolution of gas. Neither the gas nor the solid residue contained any traces of amine. Ammonia was found on treating the solidified metal with water, indicating the formation of aluminum nitride. Carbon was also found on the surface. EXAMPLE 8 Carbon disulfide decomposed rapidly upon treatment with molten aluminum, generating gaseous sulfur. The cooled melt contained both aluminum and sulfur. EXAMPLE 9 A mixed alkyl benzene sulfonate was destroyed by molten aluminum, generating a combustible gas and leaving aluminum sulfide and carbon in the metal. EXAMPLE 10 A 25 ml sample of malathion pesticide formulation (a surrogate for nerve gases VX, Soman, etc.), which contained 15 g of malathion and 9.5 g of xylene, was vaporized in the preheater and the vapors sprayed into the molten aluminum bath which was at 870.degree. C. It passed through six to twelve inches of aluminum, and the gaseous products were trapped in the water trap by displacement of water. The total vapor produced amounted to 16.6 liters. A volume of 19.2 liters was calculated to be the total volume based upon reactions which assume the total decomposition of the xylene. No detectable malathion remained in the vapors. EXAMPLE 11 A 1.8573 gram sample of arsenic trioxide is mixed with 5 g of powdered aluminum in a crucible and heated to the melting temperature of aluminum. A current of air is drawn through a funnel above the crucible and through a cold trap. Elemental arsenic is condensed in the trap and in the connecting tube. EXAMPLE 12 A 0.5055 g sample of mercuric oxide was mixed with 5 g of aluminum powder and heated to the melting point of aluminum. The resulting vapors are passed over a cold trap where they are condensed. The deposit is dissolved in nitric acid and the presence of mercury is confirmed by atomic absorption spectrophotometry. EXAMPLE 13 Osium tetroxide is contacted with molten aluminum. Osmium metal forms and dissolves in the aluminum. EXAMPLE 14 Vanadium pentoxide is contacted with molten aluminum. Vanadium metal forms and dissolves in the aluminum and alloys therewith. EXAMPLE 15 A 0.1424 g sample of freshly precipitated copper sulfide was heated with molten aluminum. The reaction was exothermic, and the copper was reduced to metals which dissolved in the molten aluminum and alloyed therewith. Aluminum sulfide was present in the slag. EXAMPLE 16 Molten aluminum was poured into a crucible containing 3.0 g sodium chromate. The chromate ion was reduced to chromium metal which dissolved in the melt, leaving no trace of oxidizing chromate ion. Sodium and aluminum oxides were left in the residue. EXAMPLE 17 A mixture of 0.5515 grams of potassium permanganate and 1.6017 grams of powdered aluminum was heated in a furnace. Vigorous exothermic reaction starts at about 600.degree. C. reducing the permanganate to manganese metal, which alloys with the aluminum. No trace of permanganate remains. EXAMPLE 18 5.0 ML of 5% sodium hypochlorite solution was mixed with 5.21 g aluminum powder in a crucible, and evaporated to dryness. When dry, it was heated to 700.degree. C. in a muffle furnace. When cooled to room temperature, the solids remaining were titrated with deionized water and filtered. No chlorine or other oxidizing agent is detected. Chloride ion is detected, indicating reduction of 99.9% of the hypochlorite ion. EXAMPLE 19 A 0.5 gram sample of sodium perchlorate was heated to 700.degree. C. in contact with powdered aluminum. Aqueous extract of the cooled solids showed the absence of any oxidizing agent, and the presence of chloride ion, indicating the complete reduction of the chlorate. EXAMPLE 20 A 0.5 gram sample of thallium nitrate is heated in a crucible with powdered aluminum while a current of air is drawn through a funnel above the crucible through a cold trap. Elemental thallium condenses in a trap and in the connecting tube. A test for nitrate was negative. EXAMPLE 21 1.1044 g of sodium sulfate was heated at 700.degree. C. in contact with powdered aluminum. The characteristic odor of hydrogen sulfide is detected in the cooled residue, indicating reduction of the sulfate. No sulfate was detected. EXAMPLE 22 1.0179 g of sodium cyanide was heated to the melting point of aluminum in powdered aluminum. Analysis of the product indicates only a negligible amount of cyanide (4.5 mg/kg). EXAMPLE 23 A hazardous waste mixture consisting of plating plant sludge from a cyanide brass plating process is dried and conveyed into molten aluminum. The cyanide is destroyed by conversion to carbon and nitrogen. Copper metal and zinc metal remain dissolved in the aluminum as harmless alloying metals. EXAMPLE 24 A disposed plastic syringe used for injecting a small volume of radioactive thallium 201 chloride into a human, for example for use in intravenous myocardial perfusion, is placed in the system as illustrated in FIG. 1 and is introduced into the molten aluminum at a temperature in excess of 710.degree. C., wherein the plastic syringe decomposes and the thallium is reduced and vaporizes and is removed by condensation. EXAMPLE 25 Inserting a plastic syringe containing a residue of strontium 85, which has been injected into a patient in a solution prepared from strontium chloride, into the system shown in FIG. 1 so that the syringe is placed in the molten aluminum at a temperature in excess of 873.degree. C. so that the plastic syringe decomposes and the strontium will form an oxide and remain as part of the slag in the aluminum pool and the resulting aluminum chloride will vaporize and be trapped in the trap system. It is understood that the invention can be practiced with any of the procedures on any halogenated wastes, whether hazardous or not; using any metals or mixtures of metals, under various conditions of temperature and pressure; including those set forth hereinabove but not limited thereto. The selection of the metals, eutectic mixtures, temperatures and apparatus can be varied. Those skilled in the art can readily vary and adapt the teachings of the invention to a set of circumstances found in a certain situation. Clearly, the method and system of the present invention are highly versatile insofar as they can handle a variety of waste mixtures including radioactive wastes in a manner whereby the contaminated materials and, in particular, radioactive materials can be controlled by virtue of the temperature of the molten reducing metal (e.g. aluminum) relative to the melting and boiling points of the radioactive elements and compounds being treated. Since certain changes may be made in the above described apparatus and method without departing from the scope of the invention involved, it is intended that all master contained in the description thereof or shown in the accompanying drawings shall be interpreted as illustrative and not in a limiting scope.
051732513	abstract	A mixing apparatus for a plurality of turbulently flowing fluid flows varying in temperature and/or composition includes a mixing chamber having a non-circular or predetermined cross section and a straight or curved center line. A plurality of single-conduit and parallel or radial deflector elements are disposed beside the mixing chamber and staggered in the direction of the center line. The deflector elements receive a fluid flow being oriented at an angle relative to the center line and staggered laterally. The deflector elements deflect the fluid flow tangentially into the mixing chamber. The mixing chamber has an outlet opening for the exit of a mixed fluid flow. In a gas-cooled, high-temperature nuclear reactor with a circular outline, the mixing apparatus has a horizontal, annular mixing chamber with a plurality of sectors. Horizontal annular conduits receive at least one fluid flow. A plurality of vertical conduits are disposed above the sectors and have upper ends connected to the horizontal annular conduits and lower ends connected to the mixing chamber. An outlet opening communicates with the mixing chamber. The sectors havie a plurality of bores formed therein in the vicinity of the outlet opening for receiving absorber material.
	summary
040452859	description	DETAILED DESCRIPTION OF PREFERRED EMBODIMENTS FIGS. 1 and 2 show an explosion-proof, pressure-tight safety vessel 1 of cylindrical shape made of reinforced concrete. Centrally inside the safety vessel are arranged a helium-coated high-temperature reactor 2 having ball-shaped fuel elements, together with the components of the primary circuit (the tubular cracking ovens, steam generators, blowers, gas lines, and the secondary cooling systems) as well as the recuperative heat-exchangers, which shall be further described below. The high-temperature reactor 2 is built into a cavity 3. Above the reactor cavity is shown a collection chamber 4 in which the cold helium is accumulated prior to being fed into the reactor. Underneath the floor of the reactor core a pillared collection chamber 5 is provided, wherein the exiting helium is accumulated after being heated in the core. The nuclear reactor 2 is connected to the primary circuit by four symmetrically and radially installed inlet and outlet pipes. In a circle around reactor cavity 3, four perpendicular pods 6, 7, 8, 9 are arranged, spaced symmetrically 90.degree. with respect to each other. Parallel thereto are arranged four additional perpendicular pods 10, 11, 12, 13, likewise spaced symmetrically to each other but on a circle of larger radius (than pods 6, 7, 8, 9) around cavity 3. These large passageways, which like the reactor cavity 3 are encased in heat-insulated and water-cooled steel liners, are capped by explosion-proof lids 15, which are secured with an excess number of fasteners. In a circle of even larger radius, four additional pods 20, 21, 22, 23 are provided, likewise positioned symmetrically at 90.degree. with respect to one another around cavity 3. These pods have a relatively small diameter and are closed on top and at the bottom by explosion-proof lids 24, 25. The pods are likewise lined with heat-insulated steel liners 14. In each of the four pods 6, 7, 8, 9, a tubular cracking oven 16 is installed on the same level as the reactor core. Each tubular cracking oven 16 is connected to a steam generator 17 installed in pods 10, 11, 12, 13. Underneath each steam generator 17 in pods 10, 11, 12, 13 a blower assembly 18 is installed comprising a single-stage axial blower, as shown in FIG. 2. In each of the pods 20, 21, 22, 23, a recuperative heat exchanger 19 is installed, each of which is connected to one tubular cracking oven 16. The tubular cracking ovens 16, the steam generators 17, and the recuperative heat exchangers 19 are accessible for removal from the top, while contrariwise, blower assemblies 18 are easily removed from below. The pods 6, 7, 8, 9 are each connected to reactor cavity 3 by a horizontal passageway 26 in which is installed coaxial gas duct 27, wherein the hot exhaust gas from the reactor streams through the inner coaxial duct 28, and the gas entering the reactor passes through the outer annular conduit 29. Then in the annular space 30 between the high-temperature reactor 2 and the reactor cavity 3 the gas passes to collection chamber 4. The tubular cracking ovens 16 are each suspended from a support plate 31 which is firmly attached by flanges to liners 14 inside each pod. The joint between the support plates 31 and the liner 14 is tightly sealed, creating the chamber 32 which is completely separated from the tubular cracking oven. Chamber 32 is charged with pure helium gas at a somewhat higher rate of pressure than that of the primary gas. Horizontal passageways 33 are installed below the support plates 31 of tubular cracking ovens 16; then connect pods 6, 7, 8, 9 with the adjacent pods 10, 11, 12, 13. In passageways 33, coaxial gas tubes 34 are also installed. The gas streams out of the tubular cracking ovens 16 in the inner tubes 35 to the steam generators 17, whereas the relatively cold gas, after being compressed by compressor 18, passes through the outer annular space 36 back to the high-temperature reactor 2. The entire primary circuit is therefore split up into four identical loops, which are combined via nuclear reactor 2 and each of which comprises a tubular cracking oven 16, a steam generator 17, a blower assembly, and the corresponding system of gas lines. Above the support plates 31 of each tubular cracking oven 16, a horizontal passageway 37 is provided which connects pods 6, 7, 8, 9 each with respective pods 20, 21, 22, 23. In each passageway 37 a coaxial duct 38 is installed comprising an inner conduit 39 and an outer annular conduit 40, through which the process gas is moved toward the tubular cracking oven 16. The hot cracking gas passes through the inner ducts 39 from the tubular cracking oven 16 to the recuperative heat exchanger 19. Four additional pods 41, 42, 43, 44 are provided which are symmetrically installed in a circle having a smaller radius than that of pods 6, 7, 8, 9. Pods 41, 42, 43, 44 are installed at 90.degree. angles and serve as receptacles for the secondary cooling system 45 which is not shown in detail. This secondary cooling system 45 is connected to nuclear reactor 2 by a radial, coaxial gas duct 46 and has the capacity of disposing of 50% of the residual fission heat. The tubes connecting the recuperative heat exchangers 19 to the components for the gasification of coal (not shown) are installed in the lids 25 of pods 20, 21, 22, 23. The methane/steam mixture passes through pipe 47 into pods 20, 21, 22, 23. It streams through annular conduit 48 into the space between the liners 14 and the coaxial ducts 51 before passing into the recuperative heat exchangers 19 through which it flows on the shell side. The mixture then passes through the exterior tubes 40 of the coaxial duct 38 to the tubular cracking ovens 16 from where it is brought to a distributor chamber 49 in each loop. Here the mixture is fed into a plurality of cracking tubes which are welded into the support plate 31. The gas, after being cracked, is collected in accumulation chamber 50 from where it passes through the interior tube 39 of the coaxial system 38 into the recuperative heat exchanger 19. The cracking gas is distributed over the tubes of the heat exchangers 19 by distributor heads 59. It flows in downward direction and is then fed through collector heads 60 into the coaxial ducts 51 installed in the lower sections of pods 20, 21, 22, 23. Coaxial ducts 51 subsequently pass out of the concrete pressure vessel 1 through lids 25 and are extended outwardly so as to connect the system with the components for the gasification of coal. For the purpose of replacing the catalyst contained in the cracking tubes, the chambers 32 above support plates 31 in pods 6, 7, 8, 9 are made easily accessible by means of access passages 52 set into pod covers 15. The pipes connecting the steam generators 17 with the steam turbine assemblies (now shown) are installed so as to pass through the lids 15 of pods 10, 11, 12, 13. The water supply passes through pipes 53 into the distributors 54 where it is distributed over a plurality of steam-generating tubes. After being vaporized and subsequently superheated in the steam generators 17, the steam is collected in collector chambers 55 from where it passes through pipes 56 to the steam turbine assembly. In the following, the circuit of the primary gas through the reactor and the circuit of the process gas through one of the tubular cracking ovens is once again summarized, limiting the description to one of the four identical loops. The cold helium gas is brought into accumulator chamber 4 above the reactor core at a temperature of 420.degree. C. and at a pressure of 39.9 bar. It then flows downwardly through the reactor core absorbing heat, and is collected in the pillared collection chamber 5. It is then distributed over the four parallel loops via the four reactor outlet pipes. The helium gas heated to 930.degree. and at a pressure of 39.2 bar enters, from below, through hot-gas duct 28 into the tubular cracking oven 16 where it is cooled down to a moderate temperature by the process gas moving through the cracking tubes in the opposite direction. The primary gas is brought through the inner coaxial tube 35 of connecting duct 34 to the steam generator 17, entering into it at a temperature of 780.degree. C. and at a pressure of 39.1 bar. It flows through it also on the shell side, but from top to bottom. In its passage through steam generator 17, the gas is cooled by the oppositely flowing water being vaporized. The gas streams out of the steam generator 17 at a temperature of 400.degree. C. and at a pressure of 38.7 bar and is then compressed by blower 18 to the highest pressure applied in the circuit, 40 bar, whereas its temperature is now 410.degree. C. The gas is returned from the blower 18 to the nuclear reactor through the exterior tubes of the coaxial pipes and on its path flows around all the components comprising the primary circuit, including the hot-gass pipes. The relatively cold gas passing out of blower 18 moves through the annular duct 57 installed in the space between the steam generator 17 and the steel liner 14 in an upward direction in its pod. The gas is brought into pods 6, 7, 8, 9 through the outer tube 36 of the coaxial duct 34 which connects each of pods 6, 7, 8, 9 to its respective counterpart, 10, 11, 12, 13. The gas then moves downwardly through the annular conduit 58 formed by the space between the tubular cracking oven 16 and the steel liner 14 in each pod, and then enters into the reactor cavity by way of the exterior annular conduit 29 in the horizontal passageway 26. Subsequently the helium gas passes through the annular space 30 between the high-temperature reactor 2 and reactor cavity 3 to accumulation chamber 4. In the recuperative heat exchangers 19, the methane/steam mixture is heated to 650.degree. C. and then brought to the tubular cracking ovens 16 at this temperature and at a pressure of 43 bar. The gas, after being cracked, emerges, having the temperature of 820.degree. C. and a pressure of 40 bar. In passing through the heat exchanger 19 the cracking gas gives off some of its heat to the methane/steam mixture and moves out of the heat exchangers at a temperature of approximately 520.degree. C. The temperature of the water supply fed into the steam generators is 170.degree. C., while the temperature of the live steam is 510.degree. C. The chamber 32 above the support plate 31 in all the tubular cracking ovens 16 is filled with pure helium at a pressure of 41 bar.
042241065	summary	This invention relates to a nuclear fuel element consisting of ceramic material and designed in the form of plates which are specially adapted to the conditions prevailing in light water reactors. The fissile ceramic material constituting the fuel elements is divided into a number of small plates or wafers each provided with a thin independent cladding of metal alloy, said wafers being aligned in a number of rows in such a manner as to occupy the entire surface of the fuel plate. A fuel element of this type was described for example in U.S. Pat. No. 4,038,135 of July 26th, 1977 and results in particularly rugged and reliable designs. This fuel element is nevertheless subject to a disadvantage in that each fuel wafer has to be covered with thin metal foil which in turn calls for industrial developments of a somewhat complex nature. Moreover, there are some specific applications of plate-type nuclear fuels in which it is particularly advantageous to utilize the fuel in the form of very thin plates and especially plates having a thickness of less than 2 mm. This is the case, for example, when it is desired to equip a pool-type reactor core with fuel plates having a U-235 enrichment of less than 8% as will be explained hereinafter. In point of fact, the techniques developed in U.S. Pat. No. 4,038,135 of July 26th, 1977 do not readily lend themselves to the production of a plate element of small thickness. The present invention is precisely directed to a fuel element of small thickness which offers the advantage of permitting easier and more rapid industrial manufacture while at the same time retaining all the advantages attached to the plate-type nuclear fuel element disclosed in U.S. Pat. No. 4,038,135. The fuel element in accordance with the present invention is distinguished by the fact that a lateral covering is provided for each fuel wafer by means of a grid of thin wires fitted within a frame and each mesh of said grid has the shape and dimensions of a fuel wafer in order to serve as a housing for one of these latter. In accordance with another important characteristic feature of the present invention, the grid whose meshes serve to house the different fuel wafers of a fuel element is made up of thin wires of a metal having low neutron-absorption characteristics, said wires being on the one hand joined together by electric welding and on the other hand diffusion-bonded to the side plates and end plates constituting the frame proper as well as to the cladding plates. In an advantageous alternative embodiment of the invention, the ends of the wires are welded to a framing wire which extends on all four sides of the grid. In accordance with the invention, the wires constituting the grid have a circular cross-section of very small diameter which is preferably less than 2 mm and the grid meshes are preferably of either rectangular or square shape. The corresponding fuel wafers of the fuel element have a thickness which is within the range of 1 to 2 mm, for example, and is most commonly of the order of 1.5 mm. In a preferred alternative embodiment of the invention, the wires of the grid are of Zircaloy, but the grid, the side plates and end plates as well as the cladding plates can also be formed either wholly or partly of aluminum or of an aluminum alloy such as AG. Composite solutions--with some parts of Zircaloy and others of aluminum or aluminum alloy--are also of interest.
	description	This Utility Application takes benefit of U.S. Provisional Patent Application 60/895,080 filed 15 Mar. 2007. 1. Field of Invention The Present invention relates to an improvement in a method and apparatus for regulating heater cycles to improve fuel efficiency. 2. Related Art The present inventor has three US Patents on regulating cycles to improve efficiency. These are: U.S. Pat. No. 5,971,284 to Hammer issued Oct. 26, 1999 for an Apparatus For Regulating Heater Cycles To Improve Forced-Air Heating System Efficiency; U.S. Pat. No. 5,960,639 to Hammer on Oct. 5, 1999 for an Apparatus for regulating compressor cycles to improve air conditioning/refrigeration unit efficiency; and U.S. Pat. No. 5,775,582, issued to Hammer on Jul. 7, 1998 for a Method and apparatus for regulating heater cycles to improve fuel efficiency All the teachings of all these patents are hereby incorporated by reference. Of these, U.S. Pat. No. 5,775,582, ('582) . . . for regulating heater cycles to improve fuel efficiency is the most relevant, and is the method and device on which the present invention improves. It is a method and apparatus for improving heating system efficiency. An electronic circuit senses a firing signal from a boiler energy value sensor such as a thermostat or pressuretrol. The circuit prevents the boiler energy value sensor from firing the burner, while the circuit senses an energy value of the outflow line at the boiler. The circuit monitors the outflow energy value and records the outflow energy value at a first time of the firing signal. The circuit then continually monitors the outflow energy until it detects an energy drop from the initial outflow energy value. The circuit responds to the energy drop by firing the burner. The invention self adaptively responds to present thermal load, reduces the number of on-off cycles, increases each burner run time while reducing total run time, improves fuel consumption, and reduces air pollution. Other improvements to the '582 invention have been made prior to this application. The Hammer device has achieved commercial success as the IntelliCon7-LCH LIGHT COMMERCIAL HYDRONIC HEATING SYSTEM ECONOMIZER. The disclosures of Hammer U.S. Pat. Nos. 5,775,582 and 5,971,284 and 5,960,639 are hereby incorporated by reference. The IntelliCon®-LCH 2 FIGS. 1A, 1B is a patented microprocessor-based fuel-saving controller 2 for light-commercial hydronic heating systems 4 and 6. It reduces fuel consumption, wear on boiler parts and burner emissions by actively managing the burner, in conjunction with the boiler operating-control 10, to properly match the boiler output to the required load. This controller indicates actual savings on a burner cycle by cycle basis and also indicates the averages of these cycles. In addition, certain parameters are programmable. All of the programmable parameters and savings values are stored in memory that will not be lost in the event of the unit being turned off or a power failure. Electric Ratings Power input: 24,115,220 VAC±10%, 5 Watts max., 50/60 Hz Control circuit input: 24,115,220 VAC±10%, 0.1 A max. Burden Relay Contact Form B, 10 A @ 220 VAC (General Purpose) Environmental Conditions For Indoor Use Maximum Altitude (2000M) Rated Ambient Temperature 32-120° F. (0-49° C.) Maximum Rh 90% non-condensing Mains Supply Voltage Fluctuations±10% Transient Over-Voltage Category (III) Pollution Degree (2) Operation After installation, setting the switch 11 on the controller 2 to the ‘ON’ position activates the control. The LCD display 12 indicates the various ‘modes’ of the device, sensed temperatures, and percent savings. The possible messages and their explanation are: STANDBY MODE The boiler is operating under its own internal operating-control, which has turned the burner off. This occurs for a period of time after the burner 8 has shut down. ECONOMIZER MODE The boiler operating-control 10 has requested the burner 8 to come on but the controller 2 has sensed that there is available heat, which can be used without burning fuel. The burner will remain off and useful heat will be delivered from the boiler's existing supply of residual heat. HEATING MODE The controller 2 has released the burner 8 to fire. HEATING/LO LIM The controller 2 has released the burner 8 to fire due to a load condition that has caused the water temperature to go below the programmed low limits. This condition may occur occasionally. If this message appears frequently, the boiler operating-control may need to be increased in 5° F. (3° C.) increments until the condition stops or the low limits may need to be adjusted (see Programming section) During normal operation, one of the above messages will be alternated with the message(s) below. HEAT TEMP=xxx° F. The measured value of the boiler outflow water temperature is displayed in ° F. (may be programmed for ° C.). DOM TEMP=xxx° F. The measured value of the domestic hot water outflow temperature is displayed in ° F. (may be programmed for ° C.). This message will only appear if the boiler supplies domestic hot water and the optional second sensor is installed (see Sensor Section of these instructions). I SAVE=xx.x % The calculated savings of the last burner cycle (I=Instantaneous). A SAVE=xx.x % The calculated average savings of all valid burner cycles since commissioning of the controller (A=Average). Note: This message will display after a minimum of 72 Hours of operation. During this time the power/fault indicator will flicker every second. ET HRS=xxxxx.x Total hours of Economizer time. (maximum=65,535.9 hours). The option to display this screen is programmable (Default=ON). RT HRS= Total hours of Burner run-time. (maximum=65,535.9 hours). The option to display this screen is programmable (Default=ON). Installation The controller 2 is electrically installed in series with the boiler operating-control 10 as shown in the wiring diagrams FIGS. 1A, 1B. It is very important that it 2 be installed, electrically, before any interlocks to ensure proper operation of the burner 8 and to eliminate any alarm or fault conditions that could be caused by the IntelliCon controller 2. AT NO TIME SHOULD ANY SAFETY CONTROLS OR CIRCUITS BE CIRCUMVENTED. Check and determine the voltages of the burner control circuit 10 and power circuit prior to installation. To ensure maximum savings, it is recommended that the Operating-control be set to a minimum of 170° F. If the setting is found to be higher than 170° F. it should NOT be adjusted. If a “circulator low-limit” or “B” type aquastat is used, the circulator low-limit should be set 5° F. (3° C.) below the HLOLIM setting. This value can be seen during power-up and is programmable. Positioning The unit 2 may be mounted on the equipment either vertically or horizontally. For readability of the display 12, the vertical position is preferred. The unit 2 should be mounted directly on the existing electric enclosure via the unit's standard ½″ electrical fitting 14 or surface mounted using the accessory mounting bracket (Not Shown). Wiring All wiring and connections must comply with Local and National Electrical Codes. The unit 2 should be wired as shown in the wiring diagrams FIGS. 1A and 1B. It is important to read all of the instructions and the NOTE on the other side of these instructions. Ensure that POWER TO THE UNIT 2 IS OFF DURING INSTALLATION and that all unused leads such as 31-34 are individually taped/insulated with tape 36. Sensors Insert the sensor wire 40, plug 41 or 42 FIG. 4 into the ‘Heating Water Sensor’ connector 43 located on the side of the unit. Mount the sensor on the boiler outflow pipe using tie-wraps 44 (see FIG. 2) or other secure method as close to the boiler as possible. Make sure that the sensor 21 makes good thermal contact with the pipe 50. Cover the sensor with a small piece of pipe insulation 52 (not provided) and secure in place as with tape 55-56 (see FIG. 3). For boilers which also supply domestic hot water through an internal coil, plug in a second sensor 62 to the ‘Domestic Water Sensor’ connector 63 and mount the sensor 62 on the domestic hot water outflow-pipe at the storage tank, if present, or at the boiler domestic water coil outlet-pipe, if no storage tank is used. Follow the same procedure to attach the sensor 21 as used above for the ‘Heating Water Sensor’ 21. This second sensor 62 should not be used if the boiler does not heat the domestic hot water. In the event that a sensor 21 or 62 fails, the controller automatically goes into bypass mode and returns full control of the burner to the boiler's operating-control 10, the ‘Power/Normal’ indicator 70 will blink, and the following message will be displayed at the LCD 12 to identify the faulty sensor: If this message appears check and replace the faulty sensor 21 or 62. Important—Read Carefully 1. Failure to follow these instructions may result in damage to the system or cause a hazardous condition. 2. Installer must be experienced, qualified, and in certain locations, licensed to work on the system that this control is being installed on. 3. After installation is complete, follow the check-out procedure as provided in these instructions to confirm proper system operation. 4. Intellidyne is not responsible for improper installation or any damages that may result from improper installation. 5. Actual wiring may differ from that shown in the diagrams FIGS. 1A and 1B. 6. Equipment may have controls not shown. 7. Because the IntelliCon can operate with different voltages for the power and control circuits, it has separate common wires 32, 34 for these circuits 4 &6 FIGS. 1A, 1B. It is necessary that these wires are connected to the proper commons or the unit will not function properly. See FIGS. 1A and 1B. Improper voltage selection may damage the unit and void the warranty. Checkout Recheck wiring one last time and make sure that the temperature sensor(s) is plugged into the proper connector(s) 43 or 63. The sensor(s) 21 (62) are only detected during power-up. Set the controller's switch 11 to ‘Off/Bypass’ and restore power to the boiler. Reset the controller's switch 11 to ‘On’. After a brief check of the electronics and displaying various parameters of the controller, the sensor(s) 21 (62) will be detected and the green ‘Power/Normal’ indicator 70 should light continuously. It is important to verify recognition of the sensors by viewing the temperature reading(s), on the display 12. If the installed sensor(s) 21 (62) are not detected, the IntelliCon controller 2 will not function properly. If the green indicator 70 is blinking or if the display 12 does not verify the installed sensor(s) 21 (62), turn the controller ‘Off’ 11 and check the sensor 21, 62 installation. After the sensor-check, depending upon the temperature of the boiler water at power-up, the controller 2 will go into one of its various modes. If the controller went into ‘STANDBY MODE’; note the operating-control 10 setting and force a burner call by temporarily adjusting the operating-control 10 higher and verifying the change of mode of the controller 2 to the ‘ECONOMIZER MODE’, ‘HEATING MODE’ or ‘HEATING/LOLIM’ mode. If the controller went in to the ‘ECONOMIZER MODE’ you can either wait for the water temperature to drop and for the controller to go into ‘HEATING MODE’ or ‘HEATING/LOLIM’, or by removing a sensor plug 41, the controller 2 will go into bypass mode, and the burner 8 should fire shortly thereafter. If, after adjusting the operating-control 10, the controller 2 went directly into ‘HEATING MODE’ or ‘HEATING/LOLIM’ the burner 8 should fire shortly thereafter. The burner 8 should run continuously until the call from the operating-control 10 is satisfied. Once satisfied, the burner should stop firing and the controller 2 should go into the ‘STANDBY MODE’. The controller 2 and burner 8 following the above sequence indicates a properly wired and functioning control. Make sure that if the operating-control 10 was previously adjusted, to return it to its' previous setting. If the burner 8 fires for a brief second then stops (even though the operating-control is calling for the burner to run) is likely caused by the Yellow 65 and Red 66 wires being reversed. If the controller does not come out of “STANDBY MODE” when the boiler's operating-control is calling for the burner to run, the unit is wired incorrectly. The likely cause in this situation is either a reversed Yellow 65 and Red 66 wire or an improperly connected ‘common’ connection 32, 34 for the control circuit 10. See the IMPORTANT note (number 7) above. Service and Troubleshooting After Installation and Checkout, the controller 2 does not require maintenance and will provide years of trouble free operation. The unit may be taken out of the circuit at any time by placing the switch 11 to the ‘Off/Bypass’ position. In this position, the unit 2 has no effect on the system and the burner 8 is controlled as it was prior to the IntelliCon controller's installation. This allows service personnel to troubleshoot or work on the system without the controller 2 intervening. If at any time the Power/Normal light 70 on the front panel blinks continuously, a sensor is not operating properly and The IntelliCon controller has automatically gone into ‘bypass mode’. If the message “TIMER FAULT” is displayed the switch should be placed into the OFF/Bypass position and service called. Programming The following parameters may be changed in the field by following these instructions. Pre-Purge time, Temperature indication in either degrees F. or C, Heating Water Low-Limit, Domestic Water Low-Limit, Maximum Economizer Hold-Off Time, Standby-Timer Override, and whether or not the Economizer Time and/or Burner Run-Time Hour accumulators are Displayed. The system may also be returned to factory default values and the Average Savings, Economizer Time, and Run-Time accumulators may be cleared. All of the default values have been carefully selected to result in the greatest savings for the broadest scope of heating system applications. Individual system requirements may require changes. Please note that all of these programmable parameters will affect the amount of savings. Prudent changes are strongly advised. It is very important that if there is any kind of a delay (more than fifteen (15) seconds), from the time that the Operating-control calls for the burner to start and the burner actually starts, that this time delay value be entered into the controller as a Pre-Purge time (e.g. actual pre-purge timer, Flue Damper interlock, etc.). If there is a delay and the correct value is not programmed into the controller, the savings calculations will be incorrect. All programming is achieved by inserting and removing a water temperature sensor plug 41 into the dom sensor connector 63, when directed to do so via the display on the controller. The sensor 21 or 62 must be connected to the cable or this will not work! You have ten (10) seconds to respond to any of the display prompts. The 10 second countdown is displayed on the controller's lcd display 12. Programming may be stopped or aborted at any time by turning the controller 2 off with switch 11. Any parameters that were changed will remain changed. Entering Configuration Mode: To enter configuration mode, the controller must be powered up without any sensors 21, 62 connected. When prompted insert a water sensor plug 41 into the DOM SENSOR connector 63. To confirm, remove the plug 41 when prompted. The unit 2 will then indicated that it has entered “Config Mode”. After a 4 second delay the display 12 will advance to the first programmable parameter (RESET DEFAULTS?). Any changes made to a programmable parameter will be confirmed by indicating “DATA SAVED” before advancing to the next parameter. RESET DEFAULTS This parameter will reset all of the programmable parameters to factory defaults. It will not clear any of the accumulators. RESET SAVINGS This parameter will clear the Average Savings accumulator. RESET This parameter will clear the Economizer Time accumulator. (Note: this Value is Accumulated Even if not being Displayed.) RESET RUN-TIME This parameter will clear the Run-Time accumulator. (Note: this Value is Accumulated Even if not being Displayed.) For all of the parameters that follow, after making a change and the “DATA SAVED” message is displayed, you will be given an additional chance to change that parameter again, before advancing to the next programmable parameter. PREPURGE=xxx SEC This parameter indicates the pre-purge time currently programmed into the controller 2 (default value=000 seconds). Next you will be prompted to change by inserting the sensor plug 41 within 10 seconds. If not inserted within the 10 seconds the controller 2 will advance to the next programmable parameter (For Degrees F. or C). If inserted you will be prompted to force a burner call, typically done by increasing the set-point of the operating-control 10, and then to remove the sensor plug 41 when the burner 8 starts. When prompted to “FORCE A HEATING CALL” the controller 2 will wait indefinitely (NO 10 second time-out) for the CALL. So it is not necessary to rush. FOR DEGREES F OR DEGREES C The controller 2 will prompt you to change to whatever value is NOT currently selected (default value=F). For example, if the parameter is currently set for degrees F., the only choice will be to change to degrees C. This setting will alter the indicated values of the next two (2) programmable parameters, and how the indicated temperatures are displayed when the controller 2 is in operation. HLOLIM=xxx HLOLIM=xxx° C. This parameter is used by the controller 2 to set the low-limit temperature for the heating water. When the heating water temperature goes below this setting, the controller 2 will no longer attempt to achieve any savings and will return control to the operating-control. To change this setting, plug in the sensor when prompted. The indicated value will be what is currently set in the controller (default=145° F./62° C.). Next the controller 2 will count up until the maximum settable value is reached (160° F./71° C.), and then will jump to the minimum settable value (90° F./32° C.). Remove the sensor when the desired value is reached. If the ‘Heating’ water temperature goes below this value while the operating-control is calling for the burner to run, the controller 2 will indicate “HEATING/LOLIM” on the display 12. DLOLIM=xxx° C. DLOLIM=xxx° F. This parameter is used by the controller 2 to set the low-limit temperature for the domestic hot water. When the domestic water temperature goes below this setting, the controller 2 will no longer attempt to achieve any savings and will return control to the operating-control 10. To change this setting, plug in the sensor when prompted. The indicated value will be what is currently set in the controller (default=115° F./46° C.). Next the controller 2 will count up until the maximum settable value is reached (150° F./66° C.), and then will jump to the minimum settable value (90° F./32° C.). Remove the sensor when the desired value is reached. If the ‘Domestic’ water temperature goes below this value while the operating-control is calling for the burner to run, the controller will indicate “HEATING/LOLIM” on the display. AX ECON=xxx MIN This feature of the controller is to limit the maximum amount of time that the controller is allowed to remain in the Economizer Mode. To change this setting, plug in the sensor when prompted. The indicated value will be what is currently set in the controller (default=30 minutes). Next the controller 2 will count up until the maximum settable value is reached (120 minutes), then “DISABLED”, and then will jump to the minimum settable value (10 minutes). Remove the sensor when the desired value is reached. If the controller goes in to the “HEATING MODE” as a result of this feature, there will be a period (“.”) appended to the word “MODE” on the display. ECON TIMER ON? ECON TIMER OFF? This parameter controls whether or not the Economizer Time accumulator is displayed. The controller 2 will prompt you to change to whatever value is NOT currently selected (default value=ON). For example, if the parameter is currently set for “ON”, the only choice will be to change to “OFF”. Note—the accumulator is active even if not displayed. RUN TIME ON? RUN TIME OFF? This parameter controls 2 whether or not the Burner Run-Time accumulator is displayed. The controller will prompt you to change to whatever value is NOT currently selected (default value=ON). For example, if the parameter is currently set for “ON”, the only choice will be to change to “OFF”. Note—the accumulator is active even if not displayed. MAX STBY=xxx MIN This feature of the controller is to limit the maximum amount of time that the controller is allowed to remain in the Standby Mode. To change this setting, plug in the sensor when prompted. The indicated value will be what is currently set in the controller (default=60 minutes). Next the controller will count up until the maximum settable value is reached (180 minutes), then “DISABLED”, and then will jump to the minimum settable value (45 minutes). Remove the sensor when the desired value is reached. After the last parameter is reached there will be a brief delay and the controller will reset. During this time the sensor(s) 21 or 21 and 62 should be reconnected or the controller 2 will attempt to go into the configuration mode again. If you don't react quickly enough, simply turn the controller off with switch 11, connect the sensor(s) and turn the controller 2 back on. The present invention: Senses outflow temperature using the sensors 21 and or 62 of FIGS. 2-3. Records temperature and time at a thermostat burner call. Uses software to calculate the following relationships. Calculates % of the reduction of fuel consumption that the installation saves. Displays the % saved on the readout.As shown in FIG. 5, The software calculates:% Savings=(RTn−RTx)/Rtnwhere Run Time Normal sans control is RTnwhere Run Time Extended by the control is RTx The presently preferred savings calculation is comprised of the following formula:% Savings=(RTn−RTx)/RTn Where: RTn=(K/(T0+T3))(T3) RTx=(K/(T0+T1+T2+T3))(T2+T3−PP)RTn=Extrapolated total burner run-time for the pseudo original burner cycleRTx=Extrapolated total burner run-time for the actual burner cycleK=constant used to normalize the data+3600PP=pre-purge timeT0 is time from burner shutdown to Thermostat call,T1 is Intellicon-induced delay of boiler firing time.T2 is Time from post delay unit call to the Recorded temperature at a thermostat burner call.T3 is the period during a burner run from Recorded temperature to burner shut-off.PP is the Pre-firing Purge time.andWhere: RTn=(K/(T0+T3))(T3) RTx=(K/(T0+T1+T2+T3))(T2+T3−PP) % Savings are displayed on the screen 12, as the screen scrolls through various factors such as sensor temperature, burner state, sensor functionality, and any other messages that the inventor may decide to communicate to user, installer, or repairman. This is described above as: I SAVE=xx.x % The calculated savings of the last burner cycle (I=Instantaneous). Average Savings The above formula calculates savings at the present time, for the present cycle, which might be called Instantaneous Savings (Lsav), or “I Save” on the readout. Another formula is used to calculate accumulated savings over an extended period of time. The equation for calculating the Average of the accumulated percentage of Savings is as follows:Average Savings=((Asav*(Ctotal−1))+Isav)/Ctotal where: Asav=Average Savings=Total of all Isav divided by (Ctotal−1) Isav=value calculated from % Savings equation Ctotal=accumulated number of cycles incremented by 1 at the end of T3 (Ctotal=Ctotal+1). Programming By Installer To enable the installer to set factors such as the Pre-firing Purge time, when the unit is turned on, if the sensors are not plugged in, a display screen on the unit will ask the installer to plug in a sensor to a sensor socket 43 or 63 on the unit. If the installer plugs in a sensor, the unit asks him to confirm by unplugging same. Then by plugging and unplugging in response to prompts from the screen, the installer can set local parameters to optimize the installation. This obviates the need for programming switches, that usually won't be used after the initial installation. Similarly, a repairman can engage a diagnostic routine, by plugging and unplugging in response to prompts from the screen. Such a routine would be similar to the routine described in above regarding the IntelliCon7-LCH LIGHT COMMERCIAL HYDRONIC HEATING SYSTEM ECONOMIZER. T0=Normal Burner Off-time. The time, in seconds, it takes for the water temperature to drop from the point when the burner was turned off by the high-limit control until the low-limit temperature is reached. The temperature reading of the heating water sensor (and the domestic sensor, if equipped) is stored when the low-limit temperature is reached (end of T0) T1=Extended Burner Off-time. The time, in seconds, it takes for the water temperature to drop from the low-limit to the calculated hold-off temperature (calculated by the IntelliCon control). The IntelliCon control is inhibiting the burner from firing at this point. T2=Extended On-time. The time, in seconds, it takes for the out-flow water temperature to go from the point when the IntelliCon control released the burner to fire until the temperature value that was stored at the end of T0 is reached. T3=Normal On-time. The time, in seconds, it takes for the out-flow water temperature to go from the temperature value stored at the end of T0 until the high-limit temperature is reached and the burner stops firing. PP=Pre-Purge Time. This time, in seconds, is the amount of time it takes from the instant that the IntelliCon control releases the burner to fire—until it actually fires. PP times longer than 15 seconds should be programmed into the controller for accuracy of the calculations. It is important to note that contained within the entire cycle (equal to T0+T1+T2+T3) is the original cycle. This is extrapolated from the entire cycle and is equal to T0+T3. Mathematically comparing these times to a constant allows an apples-to-apples comparison. The constant is necessary because, for example, a 5% savings of a short (time-wise) cycle is not equal to a 5% savings of a cycle of longer duration. So in essence the constant “K”, used below, is a value in seconds that represents a 1 hour period (3600 seconds) of time. By determining the number of cycles that could have occurred in this hour, and the resultant run-time that would occur during that hour, it is possible to determine the difference and thus the Savings. The constant also allows the averages to be added and re-averaged, whereby if the individual averages were not normalized, the adding and re-averaging of the averages would be meaningless. The calculated savings are understated. We feel quite comfortable understating as opposed to overstating. The start-up portions of the cycle are not applied to the extrapolated original cycle thus the original cycle gets the advantage of a hot heat-exchanger, thermal inertia of the heating media, and lack of any type of a delay from when the burner is released and actually fires.
046438670	claims	1. A refueling machine comprising a trolley, movable within a horizontal plane above fuel assemblies in a reactor core of a nuclear reactor facility, an outer, stationary mast fixedly mounted to said trolley and extending vertically downwardly therefrom, an inner mast coaxially mounted within said outer mast and telescopically movable therein and a gripper assembly fixedly secured to the lower end of said inner mast for attachment to said fuel assemblies for movement of said fuel assemblies into said outer mast and in and out of said reactor core, a basket-type framework surrounding the lower end of said stationary mast, a plurality of vertically mounted television cameras fixedly attached to said basket-type framework with their lenses oriented vertically downwardly, a plurality of light sources fixedly attached to said basket-type framework below said television cameras, and support cables secured to said basket-type framework for moving said basket-type framework vertically relative to said stationary mast. 2. The refueling machine as defined in claim 1, including a plurality of mirror assemblies mounted around the lower end of said stationary mast beneath a respective said television camera and inclined with respect to a horizontal plane at an angle of 45.degree.. 3. The refueling machine as defined in claim 1 wherein four television cameras are substantially equiangularly mounted around said basket-type framework, four light sources are substantially equiangularly mounted around the lower end of said basket-type framework, four mirror assemblies are substantially equiangularly mounted around the lower end of said basket-type framework and a television monitor assembly is connected to said television cameras, so that when said gripper assembly moves a fuel assembly through said outer mast, visual inspection of the entire external surface area of said fuel assembly can be made on said television monitor assembly. 4. The refueling machine as defined in claim 1 including a workholder mounted laterally off to one side of said reactor core and at a depth below the lower end of said stationary mast, and means to lower said basket-type holder away from said stationary mast and move the same into said workholder.
042232244	abstract	An improved charged-particle beam optical apparatus including a specimen holder which is mounted in at least one support member of the apparatus and vibrates when the support member is subjected to shock. The improvement of the invention comprises at least one damped supplemental oscillator coupled to the specimen holder at approximately the point of maximum vibration amplitude of the holder.
	summary
	description	1. Field This invention pertains generally to support grids for nuclear fuel rods and more particularly to a holding fixture to assist in assembly of support grids for nuclear fuel rods. 2. Related Art In most water cooled nuclear reactors, the reactor core is comprised of a large number of elongated fuel assemblies. In pressurized water nuclear reactors (PWR), these fuel assemblies typically include a plurality of fuel rods held in an organized array by a plurality of grids spaced axially along the fuel assembly length and attached to a plurality of elongated thimble tubes of the fuel assembly. The thimble tubes typically receive control rods or instrumentation therein. Top and bottom nozzles are on opposite ends of the fuel assembly and are secure to the ends of the thimble tubes that extend above and below the ends of the fuel rods. The grids, as is known in the relevant art, are used to precisely maintain the spacing and support between the fuel rods in the reactor core, provide lateral support for the fuel rods and induce mixing of the coolant. One type of conventional grid design includes a plurality of interleaved straps that together form an egg-crate configuration having a plurality of roughly square cells which individually accept the fuel rods therein. Depending upon the configuration of the thimble tubes, the thimble tubes can either be received in the cells that are sized the same as those that receive fuel rods therein, or in relatively larger thimble cells defined in the interleaved straps. The interleaved straps provide attachment points to the thimble tubes, thus enabling their positioning at spaced locations along the length of the fuel assembly. Referring to FIG. 1, a portion of an upper strap 2 and a lower strap 4 from a conventional grid design are shown. The straps 2,4 each include a plurality of slots 6. The slots 6 extend approximately half the height of the straps 2,4 and form tabs 7 beside each of the slots 6. The straps 2,4 are assembled by arranging the upper strap 2 perpendicular with respect to the lower strap 4 and sliding a slot 6 of the upper strap 2 into a corresponding slot 6 of the lower strap 4. While a portion of one upper strap 2 and one lower strap 4 are shown in FIG. 1, a conventional grid design typically includes twelve to sixteen sets of upper and lower straps 2,4. The upper and lower straps 2,4 may also include flow vanes 9 extending at an angle from the top portions of the upper and lower straps 2,4. An example of a portion of an assembled conventional grid 10 is shown in FIG. 2 and an elevational view of a fuel assembly 40 employing the grid 10 is shown in FIG. 3. The flow vanes 9 are not shown in FIGS. 2 and 3. The fuel assembly 40 is of the type used in a pressurized water reactor and basically includes a lower end structure or bottom nozzle 42 for supporting the fuel assembly on a lower core plate (not shown) in the reactor core region and a number of longitudinally extending guide thimbles or tubes 44 which project upwardly from the bottom nozzle 42. The assembly 40 further includes a plurality of grids 10. The grids 10 are axially spaced along and supported by the guide thimbles 44. Assembly 40 also includes a plurality of elongated fuel rods 36 transversely spaced and supported in an organized array by the grids 10. Also, the assembly 40 has an instrumentation tube 46 located in the center thereof and an upper end structure or nozzle 48 attached to the upper ends of the guide thimbles 44. With such an arrangement of parts, the fuel assembly 40 forms an integral unit capable of being conveniently handled without damaging the assembly of parts. As mentioned above, the fuel rods 36 and the array thereof in the assembly 40 are held in spaced relationship with one another by the grids 10 spaced along the fuel assembly length. Each fuel rod 36 includes nuclear fuel pellets 50 and the opposite ends of the rods 36 are enclosed by upper and lower end plugs 52 and 54, to hermetically seal the rod. Commonly, a plenum spring 56 is disposed between the upper end plug 52 and the pellets 50 to maintain the pellets in a tight, stacked relationship within the rod 36. The fuel pellets 50 composed of fissile material are responsible for creating the reactive power of the PWR. A liquid moderator/coolant, such as water or water-containing boron, is pumped upwardly through the fuel assemblies of the core in order to extract heat generated therein for the production of useful work. To control the fission process, a number of control rods 58 are reciprocally movable in the guide thimbles 44 located at predetermined positions in some of the fuel assemblies 40. Specifically, the top nozzle 48 has associated therewith a rod cluster control mechanism 60, having an internally threaded cylindrical member 62 with a plurality of radially extending flukes or arms 64 such that the control mechanism 60 is operable to move the control rods 58 vertically in the guide thimbles 44 to thereby control the fission process in the fuel assembly 40, all in a well-known manner. Assembling the grid 10 involves mating numerous upper and lower straps 2,4 together. However, the tight tolerances of the corresponding slots 6 in the upper and lower straps 2,4 make it difficult to properly align and mate the straps 2,4. In particular, it is difficult to automate the mating of the straps 2,4 and to mate multiple sets of straps 2,4 to each other simultaneously. As such, the assembly of the grid 10 is labor intensive, error prone and costly. It is thus desired to more efficiently assemble grids such as the conventional grid 10. In accordance with an embodiment of this concept these and other objects are satisfied by a holding fixture for assisting in assembly of a support grid for nuclear fuel rods and including a plurality of straps each having a plurality of slots extending approximately half a height of the straps and tabs formed beside or between the slots. The holding fixture includes an actuation plate, a support plate having a plurality of receiving members structured to receive therein straps of the support grid and having a plurality of cells, and a plurality of cam assemblies structured to move to deflect every other tab of the straps received in the plurality of receiving members. The cam assemblies are disposed in every other cell of the support plate. In accordance with another embodiment of this concept, these and other objects are satisfied by a holding fixture pair for assisting in assembly of a support grid for nuclear fuel rods and including a plurality of upper straps and a plurality of lower straps each having a plurality of slots extending approximately half a height of the straps and tabs formed beside or between the slots. The holding fixture pair includes an upper holding fixture and a lower holding fixture. The upper holding fixture includes an upper actuation plate, an upper support plate having a plurality of stand-offs structured to receive therein upper straps of the support grid and having a plurality of upper cells, and a plurality of upper cam assemblies structured to move to deflect every other tab of the upper straps received in the plurality of stand-offs. The upper cam assemblies are disposed in every other upper cell of the upper support plate. The lower holding fixture includes a lower actuation plate, a lower support plate having a plurality of notches structured to receive therein lower straps of the support grid and having a plurality of lower cells, and a plurality of lower cam assemblies structured to move to deflect every other tab of the lower straps received in the plurality of notches. The lower cam assemblies are disposed in every other lower cell of the lower support plate and the upper holding fixture flipped with respect to the lower holding fixture. FIG. 4A is an isometric view of a portion of a lower holding fixture 20 in accordance with an example embodiment of the disclosed concept. The lower holding fixture 20 may be used in conjunction with a similar upper holding fixture 20′ (see FIG. 4B) to assist with assembling a support grid for nuclear fuel rods, such as the grid 10 shown in FIGS. 2 and 3. Together, the lower holding fixture 20 and the upper holding fixture 20′ are considered a holding fixture pair. The lower holding fixture 20 is configured to hold the lower straps 4 of the grid 10 to receive the lower straps 4. The lower holding fixture 20 includes a plurality of cam assemblies 22 that are configured to interact with the lower straps 4 so as to deflect every other tab 7 of the lower straps 4. The lower holding fixture 20 includes a support plate 32 that includes a plurality of cells 24, and the cam assemblies 22 are disposed in every other cell 24. The support plate 32 includes receiving members, such as the notches 38 formed in the support plate 32, structured to receive therein the lower straps 4. The lower holding fixture 20 also includes an actuation plate 30. The actuation plate 30, support plate 32 and cam assemblies 22 will be described in more detail with respect to FIG. 7. In order to more clearly illustrate the disclosed concept, only a limited size lower holding fixture 20 is shown in FIG. 4A. It will be appreciated by those skilled in the art that the lower holding fixture 20 may include any number of cells 24 and cam assemblies 22 without departing from the scope of the disclosed concept and, in particular, it will be appreciated by those having ordinary skill in the art that the lower holding fixture 20 may include many more cells 24 and cam assemblies 22 than the number shown in FIG. 4A. Referring to FIG. 4B, a portion of an upper holding fixture 20′ in accordance with an example embodiment of the disclosed concept is shown. The upper holding fixture 20′ is divided into cells 24 and includes cam assemblies 22 in every other cell 24, similar to the lower holding fixture 20. However, instead of the receiving members being notches 38 structured to receive the lower straps 4, the upper holding fixture 20′ the support plate 32 of the upper holding fixture 20′ includes receiving members that are stand-offs 39 disposed at the corners of the cells 24 and extending from a surface of the support plate 32. The top portions of the upper straps 2 are not able to be received in the notches 38 of the lower holding fixture 20 because of interference from the flow vanes 9. The stand-offs 39, on the other hand, extend away from a surface of the support plate 32 and are able to receive the upper straps 2 without the flow vanes 9 abutting against the surface of the support plate 32. As such, the stand-offs 39 allow the upper straps 2 to be received by the upper holding fixture 20′. FIG. 5A is a top view of the lower holding fixture 20 and FIG. 5B is a bottom view of the upper holding fixture 20′. The lower and upper holding fixtures 20,20′ are similar and similarly include cam assemblies 22 and cells 24. Moreover, similar to the lower holding fixture 20, the upper holding fixture 20′ is configured to receive the upper straps 2 and to deflect every other tab 7 of the upper straps 2. However, as, shown in FIGS. 5A and 5B, the upper holding fixture 20′ is flipped over with respect to the lower holding fixture 20. As such, the cam assemblies 22 of the lower and upper holding fixtures 20,20′ face each other. Additionally, a cam assembly 22 of the upper holding fixture 20′ will face an empty cell 24 of the lower holding fixture 20 and a cam assembly 22 of the lower holding fixture 20 with face an empty cell 24 of the upper holding fixture 20′. Additionally, the cam assemblies 22 of the lower holding fixture 20 will deflect tabs 7 of the lower straps 4 in different directions than the cam assemblies 22 of the upper holding fixture 20′ deflect tabs 7 of the upper straps 2. For example, referring to the orientations of the lower and upper holding fixtures 20,20′ shown in FIGS. 5A and 5B, the cam assemblies 22 of the lower holding fixture 20 will deflect tabs 7 of the lower straps 2 to the left or right, while the cam assemblies 22 of the upper holding fixture 20′ will deflect tabs 7 of the upper straps 2 in upward or downward directions. FIG. 6 is an isometric view of a portion of an upper strap 2 and a lower strap 4 whose tabs 7 have been deflected by the cam assemblies 22 of the upper holding fixture 20′ and the lower holding fixture 20, respectively. As shown in FIG. 6, adjacent tabs 7 of each of the straps 2,4 are deflected in opposite directions. The deflections by the cam assemblies 22 cause the slots 6 to form V-shapes. When tabs 7 of the upper and lower straps 2,4 are deflected so that the slots 6 form V-shapes, as is shown in FIG. 6, the upper and lower straps 2,4 can still be mated together even when there is some misalignment between the upper and lower straps 2,4. In contrast, when tabs 7 of the upper and lower straps 2,4 are not deflected, as is shown in FIG. 1, the upper and lower straps 2,4 will not be able to be mated together if they are misaligned. The cam assemblies 22 of the lower and upper holding fixtures 20,20′ are additionally adjustable so as to be able to deflect tabs 7 of the upper and lower straps 2,4 as well as to release the deflection and allow the tabs 7 of the upper and lower straps 2,4 to return to their original shape shown in FIG. 1. For example, the cam assemblies 22 can be used to deflect the tabs 7 of the upper and lower straps 2,4 until the upper and lower straps 2,4 are mated to each other. Then, the cam assemblies 22 can release the deflection and the lower and upper holding fixtures 20,20′ can be removed the upper and lower straps 2,4. The result is the grid 10 shown in FIG. 2. However, by using the lower and upper holding fixtures 20,20′ to assist with assembling the grid 10, multiple straps are able to be mated to each other simultaneously and there is a tolerance for some misalignment between the upper and lower straps 2,4. Referring to FIG. 7, a cross-sectional view of a portion of the lower holding fixture 20 in accordance with an example embodiment of the disclosed concept is shown. The lower holding fixture 20 includes the actuation plate 30 and the support plate 32. A plurality of cam rods 34 are attached to the actuation plate 30 and extend in a direction substantially perpendicular away from the actuation plate 30. The support plate 32 includes a plurality of openings 37 that each allow a corresponding one of the cam rods 34 to pass therethrough. A plurality of pairs of lever members 40 are hingedly attached to the support plate 32 via hinges 43. The hinges 43 are attached to the support plate 32 in the openings 37. The cam rods 34 each include a pair of protrusions 45 that interact with the lever members 40 to cause the lever members 40 to open outwardly and deflect tabs 7 of the lower straps 4 or to close inwardly and stop deflecting tabs 7 of the lower straps 4. In more detail, the actuation plate 30 is able to move toward or away from the support plate 32. This actuation causes the cam rod 34, and the protrusions 45 on the cam rod 34 to move with respect to the lever members 40. In the position shown in FIG. 7, the protrusions 45 on the cam rod 34 are disposed so as to abut against a lower portion of the lever members 40, which causes the lever members 40 to close inwardly and not deflect tabs 7 of the lower straps 4. When the actuation plate 30 is actuated upward toward the support plate 32, the protrusions 45 of the cam rod 34 are moved upward and no longer abut against the lower portion of the lever members 40. The lower portions of the lever members 40 are thus able to move inward which causes the upper portions of the lever members 40 to move outward and deflect tabs 7 the lower straps 4. Thus, by movement of the actuation plate 30 with respect to the support plate 32, the lever members 40 can be controlled to deflect or stop deflecting tabs 7 of the lower straps 4. Together, one cam rod 34 and one pair of lever members 40 form one of the cam assemblies 22 shown in FIGS. 4A and 4B. The cam assemblies 22 may optionally also include a cover as shown in FIGS. 4A and 4B. The lower holding fixture 20 further includes notches 38 formed therein. The notches 38 are structured to receive the lower straps 4. Thus, all the lower straps 4 to be included in the grid 10 can be placed in the notches 38 of the lower holding fixture 20. The upper holding fixture 20′ includes similar notches 38 structured to receive the upper straps 2. All of the straps 2,4 of the grid 10 can thus be placed in the notches 38 of the lower and upper holding fixtures 20,20′ and can be mated together simultaneously using the lower and upper holding fixtures 20,20′. Although the lower holding fixture 20 has been described with respect to FIG. 7, the upper holding fixture 20′ also includes similar components and operates in a similar manner as the lower holding fixture 20. The difference between the upper and lower holding fixtures 20,20′ is that the upper holding fixture 20′ includes stand-offs 39, rather than notches 38, to receive the upper straps 2, as has been previously described with respect to FIG. 4B. Therefore, for economy of disclosure, a separate description of the upper holding fixture 20′ has been omitted and it will be appreciated by those having ordinary skill in the art that the cam rod 34 and pair of lever members 40 may be components of each of the cam assemblies 22 shown in the upper holding fixture 20′ of FIG. 4B. Referring to FIG. 8, a cross-sectional view of a portion of a lower holding fixture 200 in accordance with another example embodiment of the disclosed concept is shown. Although only a cross-sectional view is shown in FIG. 8, the lower holding fixture 200 is arranged similar to the lower holding fixture shown in FIGS. 4 and 5A. That is, the lower holding fixture 200 includes a support plate 320 that includes a plurality of cells 24 and cam assemblies 22 that are disposed in every other cell 24. However, rather than the cam rods 34 and lever members 40 forming the cam assemblies 22, as in the lower holding fixture 20 of FIGS. 4 and 5A, the lower holding fixture 200 of FIG. 8 includes first and second cam rods 340,350 that form the cam assemblies 22. The lower holding fixture 200 includes a first actuation plate 300, a second actuation plate 310 and a support plate 32. A plurality of first cam rods 340 are attached to the first actuation plate 300 and extend in a direction substantially perpendicular away from the first actuation plate 300. A plurality of second cam rods 350 are attached to the second actuation plate 310 and extend in a direction substantially perpendicular away from the second actuation plate 310. The support plate 320 includes a plurality of openings 360 that each allow a corresponding pair of the first and second cam rods 340,350 to pass therethrough. The first and second actuation plates 300,310 are configured to move with respect to each other, as is shown in more detail in FIG. 9. The first and second cam rods 340,350 move in conjunction with the first and second actuation plates 300,310 so that movement of the first and second actuation plates 300,310 causes the first and second cam rods 340,350 to move closer together or further apart. This movement of the first and second cam rods 340,350 can be used to move the first and second cam rods 340,350 to deflect tabs 7 of the lower straps 4. Protrusions 400,410 formed on the first and second cam rods 340,350 assist the first and second cam rods 340,350 in deflecting tabs of the lower straps 4. Similarly, movement of the first and second cam rods 340,350 can be used to stop the first and second cam rods 340,350 from deflecting tabs 7 of the lower straps 4. In more detail, the first and second actuation plates 300,310 can be moved such that the first and second cam rods 340,350 move closer to each other and do not abut against the tabs 7 of the lower straps 4. The lower holding fixture 200 further includes notches 380 formed therein. The notches 380 are structured to receive the lower straps 4. Thus, all the lower straps 4 to be included in the grid 10 can be placed in the notches 380 of the lower holding fixture 200. Although a lower holding fixture 200 has been described with respect to FIGS. 8 and 9, it will be appreciated by those having ordinary skill in the art that a corresponding upper holding fixture may be formed by replacing the notches 380 of the lower holding fixture 200 with stand-offs 39, such as those shown in FIG. 4B. Therefore, for economy of disclosure, a separate description of the upper holding fixture corresponding to the lower holding fixture 200 of FIGS. 8 and 9 has been omitted and it will be appreciated by those having ordinary skill in the art that that a corresponding upper holding fixture formed by replacing the notches 380 with stand-offs 39 may be formed as well as employed in conjunction with the lower holding fixture 200 without departing from the scope of the disclosed concept. As described herein, the various holding fixtures can be used to assist in assembling a support grid for nuclear fuel rods, such as the grid 10 shown in FIG. 2. By using the various holding fixtures described herein, the upper and lower straps 2,4 of the grid 10 can be mated together simultaneously. Additionally, the various holding fixtures can deflect tabs 7 of the upper and lower straps 2,4 so as to form V-shapes in their slots 6 and allow the upper and lower straps 2,4 to be mated together even if there is some misalignment between them. Thus, the holding fixtures described herein reduce the labor, cost and errors conventionally associated with assembling support grids for nuclear fuel rods. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
	abstract	A medical apparatus includes: a beam deflector having an electromagnet configured to provide a first magnetic field for deflecting a particle beam; and a current control configured to adjust a current of the electromagnet in correspondence with an energy level associated with an accelerator. A treatment planning method includes: defining control points in a treatment plan; setting up energy switching in one or more of the control points; and performing treatment optimization on the treatment plan based at least in part on the energy switching that is set up in the one or more of the control points. The medical apparatus also includes an energy adjuster configured to adjust the treatment energy so that the treatment energy has a first energy level when the beam output is at the first gantry angle, and a second energy level when the beam output is at the first gantry angle.
052346099	summary	BACKGROUND OF THE INVENTION The present invention relates to an X-ray permeable membrane for X-ray lithographic mask or, more particularly, to an X-ray permeable membrane for X-ray lithographic mask without defects or pinholes and having excellent smoothness of the surface and high transmissivity of visible light as well as excellent resistance against chemicals and moisture and high-energy beam irradiation. In recent years, the X-ray lithographic method is highlighted for fine patterning of semiconductor devices in place of the conventional photolithographic method. The X-ray lithographic method is performed by using an X-ray lithographic mask made of a frame-supported X-ray permeable membrane, which is required to satisfy following requirements including, for example, that: (1) the membrane is made from a material highly resistant and stable against irradiation with high-energy beams such as X-rays, high-energy electron beams, synchrotron orbital radiation, referred to as SOR hereinafter, and the like; PA0 (2) the membrane has a high transmissivity of visible light of at least 50% even when the membrane is thick enough to ensure good mechanical strengths so as to enable high-precision alignment; PA0 (3) the membrane is highly resistant against chemicals and moisture not to be affected or damaged in the manufacturing process of semiconductor devices using an etching solution, rinse and the like; PA0 (4) the membrane has a very smooth surface and is free from any defects such as pinholes; and so on. Several kinds of materials have been proposed heretofore and are used as a material of the X-ray permeable membrane for X-ray lithographic mask including, for example, boron nitride BN, silicon nitride Si.sub.3 N.sub.4, silicon carbide SiC and the like. Each of these known materials has its own advantages and disadvantages so that no quite satisfactory material is available in respect of all of the above mentioned requirements for the material. Among the above mentioned prior art materials, the most promising is silicon carbide and membranes of silicon carbide can be formed by the so-called chemical vapor-phase deposition method, referred to as the CVD method hereinafter, including thermal CVD method, plasma-induced CVD method and others, sputtering method and the like. A silicon carbide membrane formed by the thermal CVD method, in which decomposition of the reactants proceeds by means of thermal energy, has high crystallinity and is highly resistant against high-energy beam irradiation but has a disadvantage that the surface smoothness is very poor so that it can hardly be used as a membrane for the X-ray lithographic mask used in the manufacture of VLSIs which requires a patterning fineness of design rule 0.5 .mu.m or smaller unless the silicon carbide membrane as formed is subjected to improvement of the surface smoothness by taking a measure such as polishing, etching and the like resulting in a very high cost for the preparation of the X-ray lithographic mask. Although the surface smoothness and transmissivity to visible light of a silicon carbide membrane can be improved when the membrane is prepared by utilizing the plasma CVD method for the deposition of a silicon carbide film on a substrate, such a silicon carbide membrane necessarily contains a large amount of hydrogen originating in the reactant compounds, for which almost all of known reactant compounds contain hydrogen more or less, which causes a fatal defect that the membrane is poorly resistant against high-energy beam irradiation which causes removal of the hydrogen atoms out of the membrane. Further, silicon carbide membranes prepared by the sputtering method have disadvantages that the transmissivity to visible light thereof is usually low to be about 40% or lower when the membrane has a thickness of 2 .mu.m and have an amorphous structure susceptible to crystallization so that strain of the membrane is readily caused by the irradiation with high-energy beams such as SOR although the membrane is absolutely free from hydrogen and has excellent smoothness of the surface. SUMMARY OF THE INVENTION The present invention accordingly has an object to provide a novel X-ray permeable membrane for X-ray lithographic mask free from the above described problems and disadvantages in the membranes of the prior art. Thus, the X-ray permeable membrane for X-ray lithographic mask provided by the present invention is a membrane consisting of the elements of silicon, carbon and nitrogen, which is formed by the CVD method from a gaseous compound comprising the elements of silicon, carbon and nitrogen or a combination of gaseous compounds comprising, as a group, the elements of silicon, carbon and nitrogen. In particular, it is preferable that the membrane has such a chemical composition that the atomic ratio or molar ratio of silicon to the total of carbon and nitrogen is in the range from 0.05 to 2.0 and the atomic ratio of carbon to nitrogen is in the range from 0.2 to 5.0 and the chemical composition is expressed by the formula SiC.sub.x N.sub.y, in which the subscript x is a number in the range from 0.25 to 0.86 and the subscript y is a number in the range from 0.18 to 1.0. Further, the membrane is formed preferably by the thermal CVD method. In addition, it is a desirable condition that the content of hydrogen in the membrane does not exceed 1.0 atomic % even when the reactant gas or gases used in the CVD method include a hydrogen-containing compound.
	summary
052788873	summary	FIELD OF THE INVENTION The present invention relates to an apparatus and method for reducing the dosage of X-ray radiation received by both a patient and medical personnel during an X-ray fluoroscopic procedure, and more particularly to such apparatus and methods that confine full X-ray dosage to a central area, compensating for the reduced X-ray dosage in the peripheral areas by computer imaging enhancement. BACKGROUND OF THE INVENTION Interventional radiology procedures are becoming more prevalent for the detection and treatment of many diseases and injuries. Often an interventional radiology procedure involves the viewing of a catheter, or needle, as it is directed into a desired position within the body. Catheter based medical procedures are commonplace and include such medical treatments as balloon angioplasty, laser ablation, the installation of stints and many other valuable treatments. In such medical procedures the progress of the catheter is typically monitored, within a patient's body, by an X-ray fluoroscope imaging system. During a catheterization procedure, physicians and technicians need to position themselves next to the patient, in order to control the catheter. The overall X-ray exposure to such medical personnel can be higher than the X-ray exposure to the patient because medical personnel may do several X-ray fluoroscopic procedures in a single day and receive multiple dosages of X-ray radiation. For example, neuroangiographic procedures to repair an aneurysm or malformation may take as long as ten hours, during which the patient and physician are exposed to X-ray radiation much of the time. If the physician performs several such procedures a year, the physician quickly may exceed the recommended maximum dosage of radiation. The results of this potential for overexposure has been for highly trained physicians and other technical medical personnel to reduce their work load or to not wear their radiation monitor. Similarly, concern over overexposure may cause a physician to hurry a procedure, thus increasing the chances of making a mistake. One way to reduce X-ray exposure from fluoroscopy is to use various shielding techniques. Staff can be protected with lead aprons, imaging chain canopies, lead gloves, and eye shields. Patients can be protected with gonad shields, etc. Many of these techniques are not often used because they interfere with the clinical procedure in one way or another. In X-ray fluoroscopy it is well known that the dosage of the X-ray radiation is inversely proportional to the quantum noise in the viewed image. Prior art methods of X-ray dose reduction have addressed lowering the rate of dosage. For example, a nominal operational rate for X-ray fluoroscope is 30 frames/sec which may result in an exposure of approximately 10R/min skin dose. Prior art methods have attempted to reduce exposure by reducing the operational rate, for example, from 30 frames/sec. to 15 frames/sec. Such techniques have not been successful since a reduced frame rate necessitates an increased dosage rate per frame to minimize the quantum noise, the net result being no significant reduction in exposure. Other techniques for reducing the dosage of X-ray radiation include operating the fluoroscopy imaging system in a zoom mode; in other words, limiting the X-ray radiation to a small region and electronically magnifying that region to form the entire viewed image. Zoom mode imagery is not popular among some medical personnel because the zoomed image only permits a physician to view a small segment of a patient's body. Such a limited view makes it difficult for a physician to orient the placement of a catheter in a body, and prevents a physician from anticipating upcoming obstacles in the body until they appear in the zoomed image. In addition, in zoom mode, some X-ray systems increase the X-ray tube output dose such as to maintain a constant level of light output from the image intensifier. In that case, there is no dose saving to the patient. It is therefore a primary objective of the present invention to provide an apparatus and method for reducing X-ray radiation exposure to both patients and medical personnel without adversely affecting either the area of interest the X-ray fluoroscope procedure is being used to view, or the physician's ability to view the peripheral regions surrounding the area of interest. SUMMARY OF THE INVENTION The present invention relates to an apparatus and method for reducing the dosage of X-ray radiation incurred by a patient and medical personnel during a fluoroscopic procedure. During a fluoroscopic procedure X-rays are passed through a patient and are converted into a viewed image. Traditionally, the input X-ray beam is unattenuated across the entire field of view, even though it is herein recognized that, with some procedures, only a small area of the field of view actually requires high definition imaging. The present invention includes a filter member that attenuates the X-ray radiation in areas of the field of view that are not of primary interest. With the filter member in place, a physician can still visualize the entire field of view for the purposes of orientation and placement, except that now the areas in the viewed image outside the point of interest are of lower quality. By attenuating the X-ray radiation in the areas outside the point of interest, the integrated-area dosage of X-rays is greatly reduced, as is the chance of overexposure to either the patient or the physician. There is an analogy to the retina fovea mechanism of the human eye to track an object of interest. Thus the concept of the present invention is also referred to herein as an "X-ray fovea". The attenuation of the X-ray radiation in selective areas changes in the brightness of the viewed image. Thus, the areas of the viewed image created by the attenuated X-rays are amplified to match the brightness of the viewed image created by the unattenuated X-rays. To prevent a distinct division of the viewed image between the areas formed by the attenuated and unattenuated X-rays, special image processing algorithms must be used. In addition, the filter member can have a varying transparency to X-rays, such that a smooth transition is made between the various regions of the viewed image and no discernable transition line appears in the image. In addition to compensating the brightness in the peripheral area, one may also introduce temporal or spatial filtering to reduce noise. In accordance with an aspect of the invention, in an X-ray fluoroscopic apparatus for passing X-rays from an X-ray source to an X-ray detector, through a subject body; a radiation reduction device comprises: a filter member, being semi-transparent to X-rays, and having at least one aperture formed therethrough, such that X-rays passing through the at least one aperture remain unattenuated and strike the subject body in a common region; and wherein the X-rays passing through the filter member are attenuated and strike the subject body in a pattern that surrounds, and is adjacent to, the common region. In accordance with another aspect of the invention, the filter member includes a transition area surrounding the at least one aperture, the transition area having an increased transparency to X-rays as the transition area approaches the at least one aperture. In accordance with yet another aspect of the invention, the filter member is a substantially planar structure having a single aperture formed therethrough, the planar structure decreasing in thickness in the transition area such that the thickness of the planar structure is at a minimum at the edge of the aperture. In accordance with an aspect of the invention, in an X-ray fluoroscopic procedure wherein an image is produced by passing X-ray radiation through a subject body a method of reducing the dosage of X-ray radiation striking the subject body, comprises the steps of: selectively filtering the X-ray radiation such that attenuated and unattenuated X-ray radiation strike the subject body, the unattenuated X-ray radiation being confined to a predetermined common area surrounded by the attenuated X-ray radiation. In accordance with still yet another aspect of the invention, in an X-ray fluoroscopic procedure wherein a viewed image is produced, for monitoring the advancement of a medical instrument within a patient, by passing X-ray radiation through a patient; a method for reducing the dosage of X-ray radiation being exposed to the patient comprises the steps of: selectively filtering the X-ray radiation such that attenuated and unattenuated X-ray radiation pass through the patient, the unattenuated X-ray radiation being confined to a common region; calculating the size and position of the common region striking the patient; and altering the position of the common region to follow the advancement of the medical instrument, such that a point of interest on the medical instrument is viewed within the common region. In accordance with another aspect of the invention, an X-ray fluoroscopic apparatus for passing X-rays from an X-ray source arrangement through a subject body to an X-ray detector arrangement, including image processing arrangement coupled thereto; a radiation reduction device comprises: a controllable filter member arrangement being responsive to a control signal, and including a filter member being semi-transparent to X-rays and having at least one aperture formed therethrough, such that X-rays passing through the at least one aperture remain unattenuated and strike the subject body in a common region; wherein the X-rays passing through the filter member are attenuated and strike the subject body in a pattern that surrounds, and is adjacent to, the common region; and a control arrangement coupled to the X-ray source arrangement, to the image processing arrangement, and to the controllable filter member arrangement. In accordance with yet another aspect of the invention, the control arrangement provides the control signal to the controllable filter member arrangement for selectably placing the filter member in an operative mode. In accordance with still yet another aspect of the apparatus and method forming the context for the description of the invention, in an X-ray fluoroscopic apparatus for passing X-rays from an X-ray source arrangement through a subject body to an X-ray detector arrangement, including image processing arrangement including image intensifier arrangement, coupled thereto for providing an image and including a radiation reduction device comprising: a controllable filter member arrangement being responsive to a control signal, and including a filter member being semi-transparent to X-rays and having at least one aperture formed therethrough, such that X-rays passing through the at least one aperture remain unattenuated and strike the subject body in a common region; wherein the X-rays passing through the filter member are attenuated and strike the subject body in a pattern that surrounds, and is adjacent to, the common region; control arrangement coupled to the X-ray source arrangement, to the image processing arrangement, and to the controllable filter member arrangement, a method for tracking a catheter or probe, the catheter being characterized by at least some of the following characteristics: A. relatively thin, less than 2 mm, wire-like shape; PA0 B. begins in periphery of the image; PA0 C. has smooth edges; PA0 D. does not bend much; and PA0 E. X-ray dense as compared to the surround; PA0 1. morphologically processing the image to enhance an image of the catheter; PA0 2. threshold ENHANCED giving binary image, BINARY whereby the binary image contains silhouettes including a silhouette of the catheter; PA0 3. analyzing regions or blobs, in BINARY to find the image of the catheter using properties of the catheter including the catheter being thin, relatively rigid, and X-ray dense; PA0 4. finding the end of the catheter selected: PA0 5. ending. PA0 1. Entering a user-selected value for m in the equation PA0 2. obtaining a value of b necessary to give similar gray values in a center region and the peripheral region, whereby pixel values are average in annuli in central and peripheral regions giving G.sub.c and G.sub.p, respectively, deriving an appropriate value for b by computing EQU b=G.sub.c -m G.sub.p ; and PA0 3. given b and m, calculating new pixel values in said peripheral region using the equation G'.sub.p =mG.sub.p +b. PA0 1. segmenting the transition region symmetrically into a plurality of arcuate segments; PA0 2. obtaining the average overall intensity for a given arcuate segment over a range of radii, ranging from an inside boundary of the transition region to an outside boundary thereof; PA0 3. calculating an intensity profile of the given arcuate segment; and PA0 4. deriving from the intensity profile and applying to the given arcuate segment an intensity correction factor. PA0 a. calculating an average intensity of the given arcuate segment for the outside boundary of the transition region; PA0 b. calculating an average intensity of the given arcuate segment 42 for the inside boundary of the transition region; PA0 c. applying linear interpolation to the inside and outside boundary average intensity values to approximate what the average intensity ought to be for the full range of radii of the given arcuate segment; PA0 d. determining the difference between the overall average intensity of the given arcuate segment and the interpolated intensity, at a given radius to create a correct compensation factor for the given arcuate segment; PA0 e. adding the difference to the interpolated intensity and repeat for the entire range of radii; and PA0 f. repeating the foregoing for each of the arcuate segments into which the transition region 32 has been divided. PA0 1. recursively forming a gradient image of the common region; PA0 2. creating projections of the gradient image in the horizontal, vertical and two 45.degree. directions, whereby peaks occur in the data of each projection where the projections pass tangentially near the gradient image and the distance between two peaks on a single projection represents a possible diameter of the gradient image; PA0 3. matching peaks of the various projections to deduce the actual center and radius of the gradient image. PA0 4. smoothing the data to eliminate smaller, inconsequential peaks in the data; PA0 5. locating all pairs of peaks in the horizontal projection while ignoring all pairs of peaks which are separated by a distance that falls outside a range of possible diameters for the gradient image; PA0 6. repeating for the vertical projection, as well as the two projections in the 45.degree. direction, whereby the sampling distance along the two projections in the 45.degree. directions is 1/.sup.- 2 of the horizontal and vertical sampling distance; PA0 7. rescaling of the projection data relating to the two 45.degree. directions prior to computation; PA0 8. establishing for each projection a list of pairs of peaks that represent possible gradient pairs; PA0 9. for each pair of projections, identifying all pairs of peaks in one projection that are separated by the same distance as any of the pairs of peaks in the other projection, whereby six listings are obtained of pairs of peaks that represent possible gradient images; PA0 10. determining the center and inner radius for each possible gradient image 52 in the listings; PA0 11. comparing the center and inner radius on each list to find a close match on another list; PA0 12. identifying a close match as probably being the gradient image. wherein the method comprises the following steps: opening an input image, INPUT, with a flat, structuring element; PA1 subtracting the opened image from INPUT, to given ENHANCED; PA1 for each blob, make at least some of the following calculations: PA1 for each potential blob, compute a score which is a function of measures identified above; PA1 select identification of the catheter as a blob from the set of potential catheters having the highest score; PA1 the location of one endpoint of the medial axis farther away from the boundary is the end of the catheter selected; and apply a thinning algorithm to obtain a medial axis; PA2 compute length of the medial axis, MEDIAL.sub.-- LENG; PA2 compute AREA and PERIMETER. PA2 DISTANCE.sub.-- TO.sub.-- PERIPHERY.sub.-- i (i=1,2), the distance of two endpoints of the medial axis to a periphery of the image intensifier; PA2 AREA.sub.-- TO.sub.-- PERIMETER, a ratio of the area to the perimeter; PA2 PERIMETER.sub.-- TO.sub.-- MEDIAL.sub.-- LENG, a ratio of the perimeter to the length of the medial axis; PA2 BENDING, an average local curvature of the medial axis; PA2 compute a mean intensity within the blob from ENHANCED; PA2 identify potential catheters by the following properties: a. One of the DISTANCE.sub.-- TO.sub.-- PERIPHERY measure should be close to zero; PA3 b. AREA.sub.-- TO.sub.-- PERIMETER should be close to half the width of the catheter; PA3 c. PERIMETER.sub.-- TO.sub.-- MEDIAL.sub.-- LENG should be close to 2.0; PA3 d. BENDING should be small; PA3 e. mean intensity should be within a predetermined value. In accordance with still yet another aspect of the invention, in an X-ray fluoroscopic apparatus for passing X-rays from an X-ray source arrangement through a subject body to an X-ray detector arrangement, including image processing arrangement including image intensifier arrangement, coupled thereto for providing an image and including a radiation reduction device comprising: a controllable filter member arrangement being responsive to a control signal, and including a filter member being semi-transparent to X-rays and having at least one aperture formed therethrough, such that X-rays passing through the at least one aperture remain unattenuated and strike the subject body in a common region; wherein the X-rays passing through the filter member are attenuated and strike the subject body in an attenuated pattern that surrounds, and is adjacent to, the common region; control arrangement coupled to the X-ray source arrangement, to the image processing arrangement, and to the controllable filter member arrangement, a method for correcting gray-scale values in the attenuated pattern region, comprises the following steps: for corrected values in a peripheral region, G'.sub.p EQU G'.sub.p =mG.sub.p +b; In accordance with still yet another aspect of the invention, in an X-ray fluoroscopic apparatus for passing X-rays from an X-ray source arrangement through a subject body to an X-ray detector arrangement, including image processing arrangement including image intensifier arrangement, coupled thereto for providing an image and including a radiation reduction device comprising: a controllable filter member arrangement being responsive to a control signal, and including a filter member being semi-transparent to X-rays and having at least one aperture formed therethrough, such that X-rays passing through the at least one aperture remain unattenuated and strike the subject body in a common region; and wherein the X-rays passing through the filter member are attenuated and strike the subject body in an attenuated pattern that surrounds, and is adjacent to, the common region, with an annular transition region between the attenuated pattern and the common region; a method for compensating for differences in image intensity in the transition region, comprises the following steps: In accordance with a further, other aspect of the invention, steps 3 and 4 of the foregoing method comprise: In accordance with another, further aspect of the apparatus and method forming the context for the description of the invention, in an X-ray fluoroscopic apparatus for passing X-rays from an X-ray source arrangement through a subject body to an X-ray detector arrangement, including image processing arrangement coupled thereto for providing an image; a radiation reduction device comprising: a controllable filter member arrangement being responsive to a control signal, and including a filter member being semi-transparent to X-rays and having at least one aperture formed therethrough, such that X-rays passing through the at least one aperture remain unattenuated and strike the subject body in a common region; wherein the X-rays passing through the filter member are attenuated and strike the subject body in a pattern that surrounds, and is adjacent to, the common region; a method for locating an edge portion of the at least one aperture comprises the following steps: In accordance with still another, further aspect of the apparatus and method forming the context for the description of the invention, step 3 of the foregoing method for locating an edge portion comprises:
046408126	summary	This document includes two Microfiche Appendices: Appendix A, consisting of 5 fiche having 315 frames, lists an exemplary computer program for use in the preferred embodiment of the invention; and Appendix B, consisting of two fiche having 75 frames, provides a set of schematic diagrams of a preferred embodiment of the invention. This invention relates generally to nuclear reactors, and more specifically to a simulation, analysis and test apparatus and process for automated checkout of the nuclear control rod drive system. BACKGROUND OF THE INVENTION Reactivity and the resulting power level of a nuclear boiling water reactor (BWR) is adjusted by moving control rods in the nuclear core. A large reactor uses about 185 such rods. Each rod is driven in or out of the core in discrete steps by a hydraulic drive mechanism, which, in a typical installation, is controlled by four solenoid valves in a hydraulic control unit. A rod drive control system allows an operator to control these solenoid valves and, through them, the positioning of all the rods in the core. The basic function of the rod drive control system involves the operator's choice of a particular rod and the timed control of the solenoid valves to drive the rod in or out. The system returns status information from each hydraulic control unit. Due to the nature and sophistication of nuclear reactor systems and their associated electronics, comprehensive testing equipment has been developed. Typically, such testing equipment is built into the system, and requires considerable space and complexity. Numerous tests are performed by this equipment, including the status of control rod drive and instrumentation systems. These tests are impeded by several factors: first, the control rods may not normally be moved during shutdown plant conditions; second, the operator must monitor the tests at his or her control panel located some 200 to 500 feet from the valves and rods themselves; third, the response of the rods to control commands are complex signals which require sophisticated and highly expert knowledge to decipher. SUMMARY OF THE INVENTION What is provided as invention is a portable, microprocessor-based test instrument for the calibration, testing, and checkout of nuclear power reactor systems under simulated operating conditions. Included in the invention is a process using the instrument to check out reactor control instrumentation by simulating inputs, outputs and closed loop responses in real time, and to deliberately inject faulted signals to the system. The instrument is expandable to perform a variety of applications through the use of a master microprocessor bus with direct communication to interchangeable plug-in printed circuit board modules. Such applications include rod control simulation and analysis, neutron monitoring system analysis, and rod block monitoring system simulation and analysis. The portable test instrument of the instant invention is designed to electrically interconnect at any of a plurality of points within the control and monitoring systems of the reactor. The electrical network from the point of interconnection to the control rods is thereupon disabled and the test instrument, located at close proximity to the point of interconnection, receives serial control command signals just as would the disabled portion of the network, and generates appropriate acknowledgements and response signals just as would the disabled portion as if it were performing properly the command. Alternatively, the instrument can generate faulty response signals simulating improper responses to the command. The instrument is provided with a cathode ray tube (CRT) display which permits the operator to monitor the simulated control rod movements in a readily comprehensible fashion.
	claims	1. A method for manufacturing oxide fuel pellets, the method comprising:(step 1) preparing nuclear fuel powder containing uranium dioxide (UO2+x, x=0 to 0.20);(step 2) compacting the nuclear fuel powder prepared in step 1 to manufacture green pellets;(step 3) sintering the green pellets manufactured in step 2 at a temperature of about 1,200° C. to about 1,400° C. by using an atmosphere gas; and(step 4) reducing the green pellets sintered in step 3 at a temperature of about 800° C. to about 1,000° C. by using a reducing atmosphere gas. 2. The method as set forth in claim 1, wherein the nuclear fuel powder in step 1 further contains gadolinia (Gd2O3) or plutonium oxide (PuO2). 3. The method as set forth in claim 1, wherein the manufacturing of the green pellets in step 2 is performed under a compaction pressure of about 100 MPa to about 500 MPa. 4. The method as set forth in claim 1, wherein the manufacturing of the green pellets in step 2 is performed under a compaction pressure of about 150 MPa to about 450 MPa. 5. The method as set forth in claim 1, wherein the atmosphere gas in step 3 comprises at least one kind of compound selected from the group consisting of carbon dioxide, nitrogen, and argon. 6. The method as set forth in claim 1, wherein the sintering in step 3 is performed for about 2 hours to about 8 hours. 7. The method as set forth in claim 1, wherein the sintering in step 3 is performed for about 2 hours to about 5 hours. 8. The method as set forth in claim 1, wherein the atmosphere gas in step 4 comprises a hydrogen gas. 9. The method as set forth in claim 1, wherein the reducing in step 4 is performed for about 1 hour to about 5 hours. 10. The method as set forth in claim 1, wherein the sintering and reducing in steps 3 and 4 are continuously performed.
	claims	1. A method of suppressing deposition of radionuclides on components of a nuclear power plant, comprising the steps of:removing contaminants including an oxide film deposited on an inner surface of piping in the nuclear power plant by chemical decontamination;supplying a processing solution into the piping through a pipe having a pump for pressurizing the processing solution and a heating apparatus for heating the processing solution after removing the contaminants, wherein the processing solution:includes a first chemical including iron (II) ions and formic acid, a second chemical for oxidizing the iron (II) ions to iron (III) ions, and a third chemical for adjusting the pH of the processing solution, andis adjusted to a temperature within 60° C. to 100° C. by the heating apparatus,injecting the first chemical, the second chemical, and the third chemical substantially in this order into the pipe, andforming a ferrite film on the inner surface of the piping by using the processing solution,wherein the steps of removing contaminants, supplying the processing solution, injecting the first chemical, the second chemical, and the third chemical, and forming the ferrite film are carried out after the nuclear power plant is stopped and before restarting the nuclear power plant. 2. A method of suppressing deposition of radionuclides on components of a nuclear power plant according to claim 1, further comprising the step of connecting the pipe to the piping after stopping the nuclear power plant and before restarting the nuclear power plant. 3. A method of suppressing deposition of radionuclides on components of a nuclear power plant according to claim 1, further comprising the step of:decomposing reducing agent used in reductive removing included in the chemical decontamination after the reductive removing,wherein the step of forming the ferrite film occurs after decomposing the reducing agent. 4. A method of suppressing deposition of radionuclides on components of a nuclear power plant according to claim 1, further comprising the step of:decomposing reducing agent and a second pH adjusting agent used in reductive removing included in the chemical decontamination after the reductive removing,wherein the step of forming the ferrite film occurs after decomposing the reducing agent and the second pH adjusting agent. 5. A method of suppressing deposition of radionuclides on components of a nuclear power plant comprising the steps of:removing contaminants including an oxide film deposited on an inner surface of piping in the nuclear power plant by chemical decontamination;supplying a processing solution into the piping through a pipe having a pump for pressurizing the processing solution and a heating apparatus for heating the processing solution after removing the contaminants, wherein the processing solution:includes a first chemical including iron (II) ions and formic acid, a second chemical for oxidizing the iron (II) ions to iron (III) ions, and a third chemical for adjusting the pH of the processing solution,is adjusted to a temperature within 60° C. to 100° C. by the heating apparatus, andhas a pH adjusted within 5.5 to 9.0,injecting the first chemical, the second chemical, and the third chemical substantially in this order into the pipe; andforming a ferrite film on the inner surface of the piping by contacting the processing solution with the inner surface of the piping at a temperature between 60° C. and 100° C.,wherein the steps of removing contaminants, supplying the processing solution, injecting the first chemical, the second chemical, and the third chemical, and forming the ferrite film are carried out after the nuclear power plant is stopped and before restarting the nuclear power plant. 6. A method of suppressing deposition of radionuclides on components of a nuclear power plant according to claim 5, wherein the first chemical is a solution including the iron (II) ions prepared by dissolving metal iron in formic acid. 7. A method of suppressing deposition of radionuclides on components of a nuclear power plant according to claim 6, wherein the first chemical is prepared by dissolving the metal iron in the solution containing formic acid to form iron (II) ions after bubbling by inert gas or vacuum deaeration of said solution. 8. A method of suppressing deposition of radionuclides on components of a nuclear power plant according to claim 5, wherein inert gas is bubbled in at least one of a tank storing the first chemical, a tank storing the third chemical, and a surge tank storing the processing solution. 9. A method of suppressing deposition of radionuclides on components of a nuclear power plant according to claim 5, wherein the ferrite film is a magnetite film. 10. A method of suppressing deposition of radionuclides on components of a nuclear power plant according to claim 5, wherein a circulation loop for circulating the processing solution is formed by using the pipe and the piping. 11. A method of suppressing deposition of radionuclides on components of a nuclear power plant according to claim 5, wherein conducting the chemical decontamination comprises supplying liquid. 12. A method of suppressing deposition of radionuclides on components of a nuclear power plant according to claim 11, wherein the chemical decontamination includes an oxidative removing and a reductive removing, and the oxidative removing and the reductive removing are repeated. 13. A method of suppressing deposition of radionuclides on components of a nuclear power plant according to claim 5, further comprising the step of:connecting the pipe to the piping after stopping the nuclear power plant and before restarting the nuclear power plant,wherein the step of supplying the processing solution is carried out after connecting the pipe to the piping. 14. A method of suppressing deposition of radionuclides on components of a nuclear power plant according to claim 5, wherein the step of forming the ferrite film is conducted after the step of removing contaminants. 15. A method of suppressing deposition of radionuclides on components of a nuclear power plant according to claim 5, further comprising the step of connecting the pipe to the piping after stopping the nuclear power plant and before restarting the nuclear power plant.
044366957	claims	1. A method of producing useful energy and isotopes, said method comprising the steps of: introducing an assembly of nuclear materials into a large containing chamber directed toward a center of nuclear reaction; causing said assembly of nuclear materials to produce intense nuclear reactions at said center of nuclear reaction; substantially surrounding said center of nuclear reaction with a first region of liquid in the form of at least one substantially contiguous free-falling mass of liquid; and substantially surrounding said first region with a second region of spray. positioning sealing means beneath a plurality of holding means; introducing liquid into each of said holding means to form a mass; and removing the sealing means from beneath each of said holding means at sufficient speed to prevent interference with the descent of each mass to form said first region. introducing a subcritical mass into said chamber toward said center of nuclear reaction; and propelling first and second subcritical plugs into said chamber to engage said mass for assuring a more than prompt critical configuration. a containing chamber having a center of nuclear reaction; an assembly of nuclear materials; means for introducing said assembly of nuclear materials into said chamber to produce intense nuclear reactions at said center of nuclear reaction; means for producing at least one contiguous free-falling mass of slurry for substantially surrounding said point of concurrence; and spray means for producing a configuration of actinide containing slurry spray in said chamber substantially surrounding said at least one mass of slurry. a large chamber; means for introducing nuclear explosive means for descent into said chamber; means for causing an explosion of said nuclear explosive means at a predetermined point in said chamber; means for introducing a plurality of substantially contiguous free-falling liquid masses for descent into said chamber such that at the instant of nuclear explosion said nuclear explosive means is substantially surrounded by said plurality of masses; and means for introducing spray into said chamber. hollow means positioned above said chamber and having an open lower end portion; means for closing the lower end portion of said hollow means; means for introducing liquid into said hollow means; and means for withdrawing said closing means from the lower end portion of said hollow means to permit said liquid to fall into said chamber in a large liquid mass. plug means for engaging said hollow means lower end portion; arcuate leg means extending from said plug means; means for interacting with said leg means to withdraw said plug means from said hollow means lower end portion. means for engaging said arcuate leg means when plug means is engaging said hollow means lower end portion; and means for disengaging said engaging means upon withdrawal of said plug means. means extending through said hollow means wall for engaging an edge of said plug means during engagement of said hollow means lower end portion; and means for downwardly displacing said extending means for disengagement of said plug means upon withdrawal of said plug means. subcritical structure means; first and second subcritical slug means; means for dropping said subcritical structure means into said chamber toward said reaction point; and means for propelling said first and second subcritical slug means for interacting with said subcritical structure means to produce a more than prompt critical configuration at said reaction point and at said reaction time. means for propelling said first subcritical slug means from above said chamber to enter an upper portion of said opening; and means for propelling said second subcritical slug means from below said chamber to enter a lower portion of said opening. introducing a plurality of large substantially contiguous free-falling masses of liquid into the top portion of said containing chamber; having said masses free-fall so that said masses substantially surround a centroidal nuclear explosion; and applying spray within said chamber but outside of the assembly of said masses and nuclear explosion means. positioning hollow means above said chamber; releaseably sealing a lower portion of said hollow means; inserting liquid into said hollow means; and releasing said liquid for descent in said chamber in a large liquid mass. positioning sealing means beneath a holding means; introducing liquid into said holding means to form a contiguous mass of liquid; and removing the sealing means from beneath said holding means at sufficient speed to prevent interference with the descent of said mass. hollow means positioned above said chamber and having an open lower end portion; means for closing the lower end portion of said hollow means; means for introducing liquid into said hollow means; and means for withdrawing said closing means from the lower end portion of said hollow means to permit said liquid mass to fall into said chamber as a large contiguous mass of liquid. plug means for engaging said hollow means lower end portion; arcuate leg means extending from said plug means; means for interacting with said leg means to withdraw said plug means from said hollow means lower end portion. means for engaging said arcuate leg means when plug means is engaging said hollow means lower end portion; and means for disengaging said engaging means upon withdrawal of said plug means. means extending through said hollow means wall for engaging an edge of said plug means during engagement of said hollow means lower end portion; and means for downwardly displacing said extending means for disengagement of said plug means upon withdrawal of said plug means. 2. The method of claim 1 wherein said at least one mass has a mass density not substantially less than the mass density of said liquid. 3. The method of claim 1 or 2 wherein said liquid comprises a lean slurry of metallic actinide particles and other materials in molten alkalai metal. 4. The method of claim 3 wherein said other materials are selected from the group consisting of lithium, protium compounds, deuterium compounds and tritium compounds. 5. The method of claim 1 wherein the step of substantially surrounding said center of nuclear reaction comprises: 6. The method of claim 1 wherein the step of introducing an assembly of nuclear materials comprises the steps of: 7. The method of claim 6 wherein said mass has an opening therein and said first and second slugs are configured for insertion in said opening, and wherein said mass is gravitationally dropped into said chamber, said first slug is downwardly propelled for insertion into said opening with substantial velocity, and said second slug is upwardly propelled for insertion into said opening with substantial velocity substantially simultaneously with said first slug. 8. The method of claim 7 wherein each of said first and second slugs contains UH.sub.3 at cryogenic temperature and the step of propelling said first and second slugs includes regulating magnetic fields for interaction with each of said first and second slug means for controlling the velocities thereof. 9. The method of claim 1 wherein said step of causing intense nuclear reactions includes producing a nuclear explosion at said center of nuclear reaction. 10. Apparatus for producing useful energy and isotopes, said apparatus comprising: 11. The apparatus of claim 10 wherein said nuclear reactions include a nuclear explosion. 12. The apparatus of claim 10 wherein said assembly of nuclear materials comprises a subcritical mass, and first and second subcritical slug means which combine to form a more than prompt critical configuration. 13. The apparatus of claim 12 wherein said subcritical mass has a vertically aligned opening therethrough, and said first and second slug means are propelled for insertion into and collision in said opening. 14. Apparatus for producing useful energy, said apparatus comprising: 15. Apparatus as in claim 14 wherein for each of said liquid masses said liquid mass introducing means comprises: 16. Apparatus as in claim 15 wherein said closing means comprises: 17. Apparatus as in claim 16 wherein said closing means includes latching means for holding said plug means against said hollow means lower end portion. 18. Apparatus as in claim 17 wherein said latching means comprises: 19. Apparatus as in claim 17 or 18 wherein said hollow means has a wall and said latching means includes: 20. Apparatus as in claim 15 wherein said closing means comprises plug means extending diagonally from said hollow means and having a first end portion for engaging said hollow means lower end portion and a second end portion extending through a wall of said chamber. 21. Apparatus as in claim 14 including means for withdrawing said liquids from said chamber after each explosion. 22. Apparatus as in claim 21 including means for extracting useful thermal energy and debris of said nuclear explosions from said withdrawn liquids. 23. Apparatus as in claim 14 wherein said means for producing a large centroidal nuclear explosion comprises: 24. Apparatus as in claim 23 wherein said subcritical structure means has an opening therethrough which is in substantial vertical alignment at said reaction point, and said slug propelling means comprises: 25. Apparatus as in claim 14 including means for withdrawing vapors and gases from said large chamber between said explosions such that at the instant of nuclear explosion there is a very low atmospheric pressure within said chamber. 26. A method of containing large nuclear explosions in a large containing chamber seriatim, said method comprising the steps of: 27. The method of claim 26 wherein said step of introducing each of said masses comprises the steps of: 28. The method of claim 26 wherein the plurality of large substantially contiguous free-falling masses of liquid consists of a lesser plurality of inner free-falling masses and a greater plurality of outer free-falling masses. 29. The method of claim 26 wherein said centroidal nuclear explosion causes energetic interaction with said masses of liquid. 30. The method of claim 26 wherein said fluid consists of a lean slurry of actinide metals in molten alkali metal. 31. A method for producing a large substantially contiguous mass of liquid with a mass in excess of 10 tonnes that free-falls more than 10 meters into an explosion containing chamber with a volume in excess of 10,000 cubic meters and a very low atmospheric pressure, said method comprising: 32. Apparatus for producing a large substantially contiguous mass of liquid with a mass in excess of 10 tonnes for free-falling more than 10 meters into an explosion containing chamber of very low atmospheric pressure and of a volume in excess of 10,000 cubic meters, said apparatus comprising: 33. Apparatus as in claim 32 wherein said closing means comprises: 34. Apparatus as in claim 33 wherein said closing means includes latching means for holding said plug means against said hollow means lower end portion. 35. Apparatus as in claim 34 wherein said latching means comprises: 36. Apparatus as in claim 34 or 35 wherein said hollow means has a wall and said latching means includes: 37. Apparatus as in claim 33 wherein said closing means comprises plug means extending diagonally from said hollow means and having a first end portion for engaging said hollow means lower end portion and a second end portion extending through a wall of said chamber. 38. Apparatus as in claim 32 wherein a wall of said hollow means is also on its other side the wall for a different hollow means. 39. Apparatus as in claim 32 wherein said hollow means is so shaped and the means of withdrawing siad closing means so quick that the said contiguous mass of liquid substantially retains the shape of the hollow means as it falls.
	claims	1. A nuclear reactor system comprising:a fuel assembly including a fuel channel;fuel rods inside the fuel channel; anda temperature indicator inserted in between two or more of the fuel rods, the temperature indicator including,an indicating rod made of a first material;an outer housing having an upper opening, the outer housing being made of a second material and surrounding at least a portion of the indicating rod; anda rod holder attached to an inner surface of the outer housing, the rod holder being made of a third material and being configured to support the indicating rod such that a top surface of the indicating rod extends out of the upper opening of the outer housing, the third material having a lower melting point than the first and second materials,wherein,the rod holder extends in a first direction from a first portion of the inner surface to a second portion of the inner surface opposite the first portion,the indicating rod extends in a second direction from an upper surface of the rod holder out of the upper opening of the outer housing,a length of the rod holder in the first direction is greater than a thickness of the rod holder in the second direction,the indicating rod, the outer housing and the rod holder are each configured such that if the rod holder melts, the indicating rod will fall within the outer housing such that the top surface of the indicating rod is not visible above the upper opening of the outer housing, andthe rod holder is configured to melt at a temperature exceeding a threshold temperature in the fuel channel. 2. The nuclear reactor system of claim 1, wherein the outer housing has a bottom surface that is spaced away from the upper opening and spaced away from a first position on the inner surface of the outer housing at which the rod holder is attached to the outer housing such that the first position is in between the upper opening and the bottom surface of the outer housing. 3. The nuclear reactor system of claim 2, wherein a first distance between an uppermost surface of the indicating rod and the upper opening of the outer housing is less than a second distance between a lowermost surface of the indicating rod and the bottom surface of the outer housing. 4. The nuclear reactor system of claim 2, wherein a region within the outer housing in between the first position and the bottom surface of the outer housing is filled with air. 5. The nuclear reactor system of claim 1, wherein the third material is aluminum. 6. The nuclear reactor system of claim 1, wherein the outer housing is cylindrical in shape and the rod holder is a circular disc attached to the inner surface of the outer housing. 7. The nuclear reactor system of claim 6, wherein the circular disc is attached to the outer housing such that the circular disc spans an inner diameter of the outer housing. 8. The nuclear reactor system of claim 3, wherein the indicating rod, the outer housing and the rod holder are each configured such that if the rod holder melts, the indicating rod will fall within the outer housing such that the lowermost surface of the indicating rod contacts and rests upon the bottom surface of the outer housing. 9. A nuclear reactor system comprising:a fuel assembly including a fuel channel;fuel rods inside the fuel channel; anda temperature indicator inserted in between two or more of the fuel rods,the temperature indicator including,an indicating rod made of a first material,an outer housing having an upper opening, the outer housing being made of a second material and surrounding at least a portion of the indicating rod, anda rod holder attached to an inner surface of the outer housing, the rod holder being made of a third material and being configured to support the indicating rod such that a top surface of the indicating rod extends out of the upper opening of the outer housing, the third material having a lower melting point than the first and second materials,wherein,the rod holder extends in a first direction from a first portion of the inner surface to a second portion of the inner surface opposite the first portion,the indicating rod extends in a second direction from an upper surface of the rod holder out of the upper opening of the outer housing,a length of the rod holder in the first direction is greater than a thickness of the rod holder in the second direction,the indicating rod, the outer housing and the rod holder are each configured such that if the rod holder melts, the indicating rod will fall within the outer housing such that the top surface of the indicating rod is not visible above the upper opening of the outer housing, andthe rod holder is configured to melt at a temperature exceeding a threshold temperature in the fuel channel. 10. The nuclear reactor system of claim 9, wherein the outer housing has a bottom surface that is spaced away from the upper opening and spaced away from a first position on the inner surface of the outer housing at which the rod holder is attached to the outer housing such that the first position is in between the upper opening and the bottom surface of the outer housing. 11. The nuclear reactor system of claim 10, wherein a first distance between an uppermost surface of the indicating rod and the upper opening of the outer housing is less than a second distance between a lowermost surface of the indicating rod and the bottom surface of the outer housing. 12. The nuclear reactor system of claim 10, wherein a region within the outer housing in between the first position and the bottom surface of the outer housing is filled with air. 13. The nuclear reactor system of claim 9, wherein the third material is aluminum. 14. The nuclear reactor system of claim 9, wherein the outer housing is cylindrical in shape and the rod holder is a circular disc attached to the inner surface of the outer housing. 15. The nuclear reactor system of claim 14, wherein the circular disc is attached to the outer housing such that the circular disc spans an inner diameter of the outer housing. 16. The nuclear reactor system of claim 11, wherein the indicating rod, the outer housing and the rod holder are each configured such that if the rod holder melts, the indicating rod will fall within the outer housing such that the lowermost surface of the indicating rod contacts and rests upon the bottom surface of the outer housing. 17. The temperature indicator of claim 1, wherein the indicating rod, the outer housing and the rod holder are each configured such that if the rod holder melts, the indicating rod will fall within the outer housing such that an uppermost surface of the indicating rod changes from being outside the outer housing and visible above the upper opening of the outer housing before the rod holder melts to being inside the outer housing and not being visible above the upper opening of the outer housing after the rod holder melts. 18. The nuclear reactor system of claim 9, wherein the indicating rod, the outer housing and the rod holder are each configured such that if the rod holder melts, the indicating rod will fall within the outer housing such that an uppermost surface of the indicating rod changes from being outside the outer housing and visible above the upper opening of the outer housing before the rod holder melts to being inside the outer housing and not being visible above the upper opening of the outer housing after the rod holder melts. 19. A nuclear reactor system comprising:a fuel assembly including a fuel channel;fuel rods inside the fuel channel; anda temperature indicator inserted in between two or more of the fuel rods, the temperature indicator including,a tubular housing having an upper opening at one end and a bottom surface at the other end, the tubular housing being made of a first material and having a first length,an indicating rod made of a second material and having a second length, the first length of the tubular housing being greater than the second length of the indicating rod, anda rod holder, having a disc shape, attached to an inner surface of the tubular housing and spaced between the upper opening and bottom surface of the tubular housing defining an intermediate surface within the tubular housing, the rod holder being made of a third material having a lower melting point than the first and second materials, and the intermediate surface being spaced from the bottom surface of the tubular housing by a first distance,wherein the intermediate surface defined by the rod holder supports the indicating rod within the tubular housing such that an upper end of the indicating rod extends above the upper opening of the tubular housing by a second distance, the first distance being greater than the second distance,wherein the bottom surface of the tubular housing supports the indicating rod when the rod holder melts such that the upper end of the indicating rod extends below the upper opening of the tubular housing, andthe rod holder is configured to melt at a temperature exceeding a threshold temperature in the fuel channel.
050858238	claims	1. A control rod drive for positioning a control rod in a nuclear reactor vessel comprising: a shaft; a ballnut positioned over said shaft and being translatable upon rotation of said shaft; a piston disposed coaxially with said shaft and on said ballnut for positioning said control rod; means for selectively rotating said shaft in a first direction and in a second direction, opposite to said first direction; a housing surrounding a portion of said shaft; a gear fixedly joined to said shaft, said gear having a plurality of circumferentially spaced gear teeth; a latch arm pivotally joined to said housing, and having at least one latch tooth facing said gear teeth; and means for selectively positioning said latch arm in an engaged position to abut said latch tooth against a first one of said gear teeth for preventing rotation of said shaft in said first direction, and in a disengaged position to spaced said latch tooth away from said gear teeth for allowing said shaft to rotate without obstruction between said gear teeth and latch tooth for translating said ballnut and in turn said piston for inserting and withdrawing said control rod in said reactor vessel. a first end including said latch tooth; an intermediate portion pivotally joined to said housing; and a second end, opposite to said first end, joined to said positioning means. each of said gear teeth has a first contact surface and a second, opposite, contact surface; said latch tooth has a first contact surface and a second, opposite, contact surface; and said latch tooth and said gear teeth being complementary in configuration so that said first gear tooth first contact surface abuts said latch tooth first contact surface in said latch arm engaged position. said latch arm including an aft surface at said first end; a stop member fixedly joined to said housing and positioned adjacent to said latch arm aft surface so that rotation of said shaft in said first direction is opposed by wedging of said latch arm aft surface against said stop in said latch arm engaged position. 2. A control rod drive according to claim 1 wherein said positioning means are effective for allowing rotation of said shaft in said second direction by said rotating means to intermittently disengage said latch tooth from said gear teeth while said latch arm is in said engaged position. 3. A control rod drive according to claim 2 wherein said positioning means include an electromagnet having a solenoid, plunger, and spring, said solenoid being fixedly joined to said housing, said plunger extending from said solenoid and pivotally joined to said latch arm, and said spring being disposed in said solenoid to resiliently bias said plunger to an extended position when said solenoid is deenergized for positioning said latch arm in said engaged position, and said solenoid being energizable to draw said plunger to a withdrawn position for positioning said latch arm in said disengaged position. 4. A control rod drive according to claim 2 wherein said latch arm includes: 5. A control rod assembly according to claim 4 wherein said latch arm is arcuate and includes a concave inner surface facing said gear, and said latch tooth is disposed on said inner surface. 6. A control rod assembly according to claim 5 wherein: 7. A control rod drive according to claim 6 wherein said latch arm intermediate portion and said latch tooth are positioned relative to said first gear tooth so that said latch tooth first contact surface opposes rotation of said shaft in said first direction when said latch arm is in said engaged position due to tensile loads imposed by said first gear tooth on said latch tooth affecting elastic extension of said latch arm. 8. A control rod drive according to claim 6 wherein said latch tooth second contact surface is positioned relative to a second contact surface of a second gear tooth disposed adjacent to said first gear tooth in said latch arm engaged position so that rotation of said shaft in said second direction by said rotating means causes said second gear tooth second contact surface to push said latch tooth radially away from said gear to intermittently disengage said latch tooth from said gear teeth. 9. A control rod drive according to claim 6 further including: 10. A control rod drive according to claim 6 wherein said arcuate latch arm includes a convex outer surface having a stop portion at said latch arm second end disposed adjacent to said housing in said latch arm engaged position for contacting said housing upon predetermined wear of said latch arm intermediate portion for maintaining engagement between said latch tooth and said first gear tooth.
	summary
045047391	claims	1. A method of filling a radiation shield comprising a free-standing container formed of thin flexible material, said method comprising the steps of: first, filling the radiation shield operable to be free-standing with a gas at sufficient pressure to form the shield into its designed dimensional free-standing configuration; then second, placing the radiation shield in the desired relationship to a radiation source; and then third, replacing the gas with a radiation attenuating liquid which also forms the shield into its designed dimensional free-standing configuration. a step of filling said radiation shield with air. replacing the gas with a hydrogenous radiation attenuating liquid which is denser than water. (a) a container being operable to be free standing formed of thin flexible material; and (b) a first means for filling said container with a gas so as to cause said container to assume its designed dimensional configuration; (c) a second means for emptying the gas from said container; (d) a third means for filling said container with a radiation attenuating liquid at the same time that said second means is emptying the gas from said container so as to cause said container to maintain its designed dimensional configuration as the radiation attenuating liquid replaces the gas; and (e) a fourth means for emptying the radiation attenuating liquid from said container. (a) a container formed of thin flexible material; (b) a first means for filling said container with a gas so as to cause said container to assume its designed dimensional configuration; (c) a second means for filling said container with a radiation attenuating liquid; (d) a third means for venting the displaced gas at the same time that said second means is filling said container so as to cause said container to maintain its designed dimensional configuration as the radiation attenuating liquid replaces the gas; and (e) a fourth means for emptying the radiation attenuating liquid from said container. placing the radiation shield in the desired relationship to a radiation source; filling the radiation shield with a gas of a sufficient pressure to form the shield into its designed free-standing dimensional configuration; then replacing the gas with a radiation attenuating liquid which also maintains the shield in its original dimensional configuration. 2. A method as recited in claim 1 wherein said step of filling the radiation shield with a gas comprises 3. A method as recited in claim 1 or 2 wherein said step of replacing the gas comprises the step of: 4. A method of solidifying in its designed dimensional configuration a radiation shield comprising a container formed of thin flexible material filled with a radiation attenuating liquid, said method comprising the step of maintaining an over-pressure of gas above the radiation attenuating liquid in the shield so as to maintain the shield in its designed dimensional configuration. 5. A radiation shield for use in installations containing sources of radiation, said radiation shield comprising: 6. A radiation shield for use in installations containing sources of radiation, said radiation shield comprising: 7. A method of filling a radiation shield comprising a free-standing container formed of thin flexible materials, said method comprising the steps of:
050826174	abstract	An isotopic heat source is formed using stacks of thin individual layers of a refractory isotopic fuel, preferably thulium oxide, alternating with layers of a low atomic weight diluent, preferably graphite. The graphite serves several functions: to act as a moderator during neutron irradiation, to minimize bremsstrahlung radiation, and to facilitate heat transfer. The fuel stacks are inserted into a heat block, which is encased in a sealed, insulated and shielded structural container. Heat pipes are inserted in the heat block and contain a working fluid. The heat pipe working fluid transfers heat from the heat block to a heat exchanger for power conversion. Single phase gas pressure controls the flow of the working fluid for maximum heat exchange and to provide passive cooling.
	summary
	claims	1. A device for scanning a beam as a periodic function of time, the device comprising: a. a wheel having an axis of rotational symmetry, the wheel being opaque to a specified energy range of electromagnetic radiation, the wheel having a set of apertures for transmitting the radiation in such a manner that the radiation is emitted in a beam from each of a specified number of illuminated apertures at a time; b. a first rotary actuator coupled to the wheel for rotating the wheel about a rotation axis coincident with the axis of rotational symmetry of the wheel such that the beam is scanned in a plane perpendicular to the axis of rotational symmetry of the wheel; and c. a second rotary actuator for rotating the wheel about a scan axis not parallel to the axis of rotational symmetry of the wheel. 2. A device in accordance with claim 1 , wherein at least one of the first and second rotary actuators is a motor. claim 1 3. A device in accordance with claim 1 , wherein the scan axis is perpendicular to the rotation axis. claim 1 4. A device in accordance with claim 1 , wherein the rotation about the rotation axis is faster than the rotation about the scan axis. claim 1 5. A device in accordance with claim 1 , wherein the rotation about the scan axis subtends less than a full circular rotation. claim 1 6. A device in accordance with claim 1 , further comprising a source of electromagnetic radiation for emitting radiation incident upon an inner surface of the wheel. claim 1 7. A device in accordance with claim 6 , wherein the source of electromagnetic radiation is fixed relative to at least one of the rotation axis and the scan axis. claim 6 8. A device in accordance with claim 6 , wherein the source of electromagnetic radiation is disposed at a specified offset from the rotation axis. claim 6 9. A device in accordance with claim 6 , wherein the source of electromagnetic radiation is a x-ray tube that emits x-rays. claim 6 10. A device in accordance with claim 1 , wherein the wheel includes lead. claim 1 11. A device in accordance with claim 1 , wherein the wheel includes a drum. claim 1 12. A device in accordance with claim 1 , wherein the wheel includes a hoop. claim 1 13. A device in accordance with claim 1 , wherein the specified number of illuminated apertures is at least one. claim 1 14. A device in accordance with claim 1 , wherein the set of apertures includes filters to attenuate a specified range of electromagnetic energies. claim 1 15. A device in accordance with claim 1 , wherein the set of apertures is regularly spaced. claim 1 16. A device in accordance with claim 1 , wherein at least one aperture from the set of apertures differs from at least one other aperture from the set of apertures in at least one characteristic selected from the group of size, shape, and transmission spectrum. claim 1 17. A method for scanning a beam in two dimensions as a periodic function of time, the method comprising: a. illuminating a surface of a wheel with electromagnetic radiation to which the wheel is opaque other than at a set of apertures traversing the wheel in a direction of propagation of the electromagnetic radiation; b. rotating the wheel about an axis of rotational symmetry of the wheel; and c. simultaneously rotating the wheel about an axis not parallel to the axis of rotational symmetry of the wheel. 18. A method in accordance with claim 17 , wherein the orientation of the beam is scanned in a plane parallel to an axis of rotational symmetry of the wheel. claim 17
	abstract	A system for electron pattern imaging includes: a device for converting electron patterns into visible light provided to receive an electron backscatter diffraction (EBSD) pattern from a sample and convert the EBSD pattern to a corresponding light pattern; a first optical system positioned downstream from the device for converting electron patterns into visible light for focusing the light pattern produced by the device for converting electron patterns into visible light; a camera positioned downstream from the first optical system for obtaining an image of the light pattern; an image intensifier positioned between the device for converting electron patterns into visible light and the camera for amplifying the light pattern produced by the device for converting electron patterns into visible light; and a device positioned within the system for protecting the image intensifier from harmful light.
	description	Referring now to FIG. 1 of the drawings, an ion beam implanter is shown generally at 10. The ion implanter includes an ion source 12, for producing a generally positively charged ion beam 14 that is extracted therefrom by known means, for example, an extraction electrode. A mass analysis magnet 16 mass analyzes the extracted ion beam 14 and outputs a mass analyzed ion beam 18 which includes only those ions having a charge-to-mass ratio that falls within a prescribed range. The mass analyzed ion beam 18 passes through a resolving aperture 20 and is implanted into wafers W situated upon pedestals situated about the periphery of a rotating support or disk 22. The rotating disk in the disclosed embodiment is made of aluminum, although it may be coated with a layer of silicon. In the case of an aluminum rotating disk 22, the disk would be more electrically conductive than the wafers situated thereon. In the case of a silicon-coated disk 22, the disk would generally be less electrically conductive than the wafers situated thereon (dependent upon whether or not a patterned insulating surface such as a photoresist is applied to the wafers). Generally, the invention acknowledges that the electrical conductivity of the wafers and the portions of the disk that surround them are different. This difference in electrical conductivity may be used to determine whether or not the wafer charge accumulation is adversely affecting the beam passing thereover. The disk 22 is vertically translated along an axis Y by means of a motor 24 and leadscrew 26. The disk 22 is rotated by means of motor 28, in a direction indicted by arrow 29, about an axis that passes through disk center 31 perpendicularly to the plane of the disk. The wafers W are positioned about the periphery of the disk 22 at locations that are substantially equidistant from the disk center. The full surface area of the wafers W are implanted as they rotate in a circular path (in the xe2x80x9cX scanxe2x80x9d direction) and are vertically translated (xe2x80x9cin the Y scan directionxe2x80x9d) before the fixed position ion beam 18. Ion dosage received by the wafers W is determined by rotational velocity and the vertical translational velocity of the spinning disk 22, both of which are determined by the motor control 30. Charge neutralization system 33 is provided for neutralizing the positive charge that would otherwise accumulate on the wafers as they are implanted by the generally positively charged ion beam 18. U.S. Pat. No. 5,959,305, which discloses a known type of charge neutralization system, is hereby incorporated by reference as if fully set forth herein. The present invention is embodied as an in-process charge monitor and control system 32. The system 32 includes means to measure the amount of charge accumulation on the wafers W that can cause the ion beam to change shape as the disk rotates, causing the beam to successively pass from the wafers to the intermediate portions of the conductive aluminum disk surface. In response to these measurements, the operation of the charge neutralization system 33 may be adjusted or tuned, as further explained below. Alternatively, the output of system 32 may be used as and input to a dose control system 35 to control the rotation and translation of the spinning disk 22 to insure a uniform implant dose across the entire surface of the wafers W being implanted. The dose control system 35 includes known elements such as a Faraday cage 34 providing an output signal 36. The output 36 from the Faraday cage 34 and an output 41 from a pressure monitor disposed within the implantation chamber, such as an ion gauge 43, are input to the control circuitry 50. The circuitry 50 uses these inputs to determine an appropriate X-scan and Y-scan speed of the wafer in front of the ion beam 18, as is known in the art. Specifically, the Faraday cage 34 is mounted behind the spinning disk 22 and is used to measure the ion beam current that passes through slot 62 in the disk. The length of the slot 62 is at least as long as the diameter of the wafers being implanted (e.g., 200 mm or 300 mm) so that the slot will receive ion beam current throughout the entire range of the Y-scan of the wafers (see also FIG. 3). The dose control circuitry 50 outputs control signal 52 to motor control 30 based on the outputs of the Faraday cage 34 and the ion gauge 43. Motor control 30 in turn outputs rotational control signal 54 to motor 28 and vertical translational signal 56 to motor 24, in order to maintain a uniform implantation across the surface of the wafers being implanted. In this manner, the outputs of Faraday cage 34 and the ion gauge 43 are used by the control circuitry 50 to thereby determine the dose of ions implanted into the wafers. The control circuitry also includes memory 58 and a user console or interface 60. The use of the output of the Faraday cage 34 and ion gauge 43 to control rotational and translation movement of the wafers W in front of the ion beam 18 is known. However, using only these mechanisms may result in non-uniform wafer implants because the ion beam current measurement provided by Faraday cage 34 does not take into account changes or disturbances to the ion beam profile as it passes from portions of the conductive aluminum disk surface intermediate wafers, to the insulative charged surface of a particular wafer being implanted. For example, the ion beam may xe2x80x9cblow-upxe2x80x9d, or become less controllably focused, if it is exposed to a sufficiently positive charge accumulation over the wafer being implanted. As such, a non-uniform wafer implant may be obtained. FIG. 2 shows one example of such a non-uniform implant, commonly referred to as a xe2x80x9cbull""s-eyexe2x80x9d pattern of non-uniform ion implantation. As shown in FIG. 2, the areas of the implanted wafer marked with xe2x80x9c+xe2x80x9d indicate areas of overdose (low sheet resistivity), and the areas marked with xe2x80x9cxe2x88x92xe2x80x9d indicate areas of underdose (high sheet resistivity). FIG. 2 resulted from implanting a 200-mm wafer with boron (B) ions at an energy level of 2 kilo-electron-volts (keV). As such, the present invention provides an additional ion beam measurement mechanism that takes into account changes or disturbances in ion beam profile, in order to improve dose uniformity across the surface of the wafer. Referring back to FIGS. 1 and 3, the in-process charge monitor and control system 32 includes electrical charge pick-ups or monitors 38 and 40 for outputting signals 42 and 44, respectively, and a comparator 46 for comparing the signals 42 and 44. Apertures 64 and 66 are provided in the disk 22 to receive portions of the ion beam current when it passes thereover. As shown in FIG. 1, portions of the ion beam are shown in phantom as reference numerals 18a and 18b as indicative of the portions of the beam that will pass through aperture 64 and 66, respectively, when the disk 22 rotates from the position shown in FIG. 1. Aperture 64 and aperture 66 are located the same distance d from disk center 31. As the disk 22 rotates, a first portion 18a of the ion beam current passes through aperture 64 and is measured by charge pick-up or monitor 40, which produces output signal 44. As the disk 22 continues to rotate, a second portion 18b of the ion beam current passes through aperture 66 and is measured by charge pick-up or monitor 38, which produces output signal 42. Aperture 64 is selected at a location where the ion beam is unaffected by the charge accumulation on the wafer, and aperture 66 is selected at a location where the ion beam is affected by the charge accumulation on the wafer. In other words, aperture 66 is located closer to a wafer than is aperture 64. Alternatively, the first and second apertures may each be located equidistant from a wafer but surrounded by portions of the disk having different electrical conductivity characteristics. For example, aperture 64 may be provided in a portion of the disk that is aluminum, and aperture 66 may be provided in a portion that is silicon coated. In either case, comparator 46 compares the output signals of charge monitors 38 and 40 to determine the effect, if any, that the charged insulative surfaces of the wafers have on the ion beam profile. For example, in the disclosed embodiment of FIG. 3, if the comparator 46 detects no measurable difference in the first and second portions of the beam current, it can be determined that there is no adverse effect causing beam xe2x80x9cblow-upxe2x80x9d. The negligible comparator output 48 indicates that the charge neutralization system 33 of the ion implanter is operating to effectively neutralize any charge accumulation on the wafers and permit a uniformly dosed implant. However, if the comparator 46 detects a measurable difference in the first and second portions of the beam current, it can be determined that there is an adverse effect causing beam xe2x80x9cblow-upxe2x80x9d. For example, if the beam is xe2x80x9cblown-upxe2x80x9d, the peak ion beam current measured at aperture 66 would be less than that measured at aperture 64. Alternatively, one can measure the time distribution of the beam as it passes apertures 64 and 66. If the beam is detected for a longer period of time at aperture 66, it indicates a beam xe2x80x9cblow-upxe2x80x9d condition. In either case, the measurable comparator output 48 indicates that the charge neutralization system 33 of the ion implanter is not operating to effectively neutralize any charge accumulation on the wafers and permit a uniformly dosed implant. As such, the operation of the charge control system (33) may be adjusted or tuned, using comparator output 48, to provide a greater supply of low energy electrons for neutralizing this excess wafer charge accumulation. Alternatively, the output 48 of comparator 46 may be used instead to adjust the dose control circuitry 50. (As shown in FIG. 1, comparator output 48 is shown in phantom as an alternative input to dose control circuitry 50.) For example, the bull""s-eye pattern of FIG. 2 may be correlated to the output 48 of comparator 46. As such, the dosage control circuitry 50 may be programmed to adjust the X-scan and Y-scan speeds of the disk in real time to correct for the anticipated dosage errors. In effect, the dosage control circuitry 50 uses comparator output 48, in addition to the outputs of the ion gauge 43 and the Faraday cage 34, to modify its output control signal 52 to motor control 30. However, it is anticipated that the invention may be more directly implemented as a means to tune the operation of the charge neutralization system 33, as described above. Accordingly, a preferred embodiment of an in-process charge monitor and control system has been described. With the foregoing description in mind, however, it is understood that this description is made only by way of example, that the invention is not limited to the particular embodiments described herein, and that various rearrangements, modifications, and substitutions may be implemented with respect to the foregoing description without departing from the scope of the invention as defined by the following claims and their equivalents.
047042354	claims	1. A method of decontaminating radionuclide-contaminated acid-insoluble corrosion products from primary system surfaces in pressurized water reactors comprising contacting the contaminated surfaces with an oxidation agent in an acidic solution and dissolving the corrosion products which have been made acid-soluble by the oxidation, the oxidation being performed with a water-based oxidation agent having a pH below 7 and containing 0.01-50 g/l cerium nitrate, 0.01-50 g/l chromic acid and 0.001-1 g/l ozone. 2. A method according to claim 1 wherein the oxidation agent is an acidic aqueous solution of cerium nitrate and chromic acid and ozone in a saturated solution and dispersed form. 3. A method according to claim 1 wherein the oxidation agent is a two-phase ozone gas-aqueous mixture, where ozone in gaseous form has been dispersed in an acidic aqueous solution of cerium nitrate and chromic acid. 4. A method according to claim 1 wherein the oxidation and dissolution are performed in one and the same step. 5. A method according to claim 1 wherein the oxidation and dissolution are carried out at a temperature below about 60.degree. C. 6. A method according to claim 1 wherein the water-based oxidation agent has been made acidic with nitric acid. 7. A method according to claim 6 wherein the water-based oxidation agent has been made acidic to a pH value of about 1. 8. A method according to claim 1 wherein the concentration of cerium nitrate is 0.5-2 g/l, the concentration of chromic acid is 0.05-0.2 g/l, and the concentration of the ozone is 0.005-0.015 g/l. 9. A method according to claim 2 wherein the oxidation and dissolution are performed in one and the same step. 10. A method according to claim 3 wherein the oxidation and dissolution are performed in one and the same step. 11. A method according to claim 2 wherein the oxidation and dissolution are carried out at a temperature below about 60.degree. C. 12. A method according to claim 3 wherein the oxidation and dissolution are carried out at a temperature below about 60.degree. C. 13. A method according to claim 4 wherein the oxidation and dissolution are carried out at a temperature below about 60.degree. C. 14. A method according to claim 9 wherein the oxidation and dissolution are carried out at a temperature below about 60.degree. C. 15. A method according to claim 2 wherein the water-based oxidation agent has been made acidic with nitric acid. 16. A method according to claim 3 wherein the water-based oxidation agent has been made acidic with nitric acid. 17. A method according to claim 4 wherein the water-based oxidation agent has been made acidic with nitric acid. 18. A method according to claim 2 wherein the concentration of cerium nitrate is 0.5-2 g/l, the concentration of chromic acid is 0.05-02 g/l, and the concentration of ozone is 0.005-0.015 g/l. 19. A method according to claim 3 wherein the concentration of cerium nitrate is 0.5-2 g/l, the concentration of chromic acid is 0.05-02 g/l, and the concentration of ozone is 0.005-0.015 g/l. 20. A method according to claim 4 wherein the concentration of cerium nitrate is 0.5-2 g/l, the concentration of chromic acid is 0.05-0.2 g/l, and the concentration of ozone is 0.005-0.015 g/l. 21. A method according to claim 5 wherein the oxidation and dissolution are carried out at a temperature below about 25.degree. C. 22. A method according to claim 21 wherein the temperature is below about 20.degree. C. 23. A method according to claim 11 wherein the oxidation and dissolution are carried out at a temperature below about 25.degree. C. 24. A method according to claim 23 wherein the temperature is below about 20.degree. C. 25. A method according to claim 12 wherein the oxidation and dissolution are carried out at a temperature below about 25.degree. C. 26. A method according to claim 25 wherein the temperature is below about 20.degree. C. 27. A method according to claim 13 wherein the oxidation and dissolution are carried out at a temperature below about 25.degree. C. 28. A method according to claim 27 wherein the temperature is below about 20.degree. C. 29. A method according to claim 14 wherein the oxidation and dissolution are carried out at a temperature below about 25.degree. C. 30. A method according to claim 29 wherein the temperature is below about 20.degree. C.
053533143	description	DETAILED DESCRIPTION OF THE INVENTION The following description is of the best mode presently contemplated for carrying out the invention. This description is not to be taken in a limiting sense, but is made merely for the purpose of describing the general principles of the invention. The scope of the invention should be determined with reference to the claims. Referring first to FIG. 1, there is shown a diagrammatic view of the main elements of a tokamak 20, with a portion thereof cutaway. The design and operation of such a tokamak is well described in the art, see, e.g., Artsimovich and/or Furth, supra. Only a very cursory overview of the tokamak's construction and operation is thus presented herein. Basically, the tokamak 20 includes a toroidal vacuum vessel 22 that is centered about a major axis 24. A minor axis 25, centrally located within the toroidal vessel 22, encircles the major axis 24. The relationship of the major and minor axes 24 and 25 is best seen in FIG. 1A. The vessel 22 is made from a conductive material, such as non-magnetic stainless steel or inconel, and is constructed with sufficiently thick walls to withstand the vacuum pressures that are developed therein. A large number of toroidal field magnetic coils 26 are equally spaced around the vessel 22, each encircling the minor axis 25 and a respective segment of the vessel 22. Eighteen such coils 26 are illustrated in FIG. 1, but this number is only exemplary. When energized with an electrical current, the toroidal coils 26 combine to produce a toroidal magnetic field B.sub.T, represented by the arrow 28, that encircles the major axis 24 within the vacuum vessel 22. A plurality of poloidal field magnetic coils 30 are positioned inside of the toroidal field coils 26, yet still outside of the vacuum vessel 22, so as to encircle the major axis 24. As depicted in FIG. 1, the windings of the poloidal field coils 30 are substantially perpendicular to the windings of the toroidal field coils 26. When energized with an appropriate electrical current, the poloidal field magnetic coils 30 combine to produce a poloidal magnetic field B.sub.P, represented by the arrow 32, that encircles the minor axis 25 of the vacuum vessel 22. Ohmic heating primary windings 34 are positioned inside of the toroidal field coils 26, in close contact with the vacuum vessel 22, so as to encircle the primary axis 24, much like the poloidal field coils 30. When energized with an electrical current, the field coils 30 (acting as a transformer primary winding) induce an electrical current, I.sub.p, 36 in the plasma (acting as a transformer secondary winding) to heat the plasma. Not shown in FIG. 1, but understood to be part of any tokamak or similar plasma-confining structure are conventional means for establishing a desired vacuum pressure within the vessel 22, and means for injecting the appropriate gases into the vessel from which plasma may be formed. Because plasma is an ionized gas, it is also an electrical conductor, with the movement of electrons (negatively charged particles) in one direction and the movement of positively charged ions in the other direction representing the flow of electrical current. An important part of the operation of a tokamak is the creation of axial current flow through the plasma contained within the vessel 22. Such current flow follows the minor axis 25 and is depicted in FIG. 1 by the arrow 36. In operation, appropriate gases are introduced into the vacuum vessel at the appropriate pressure. These gases, e.g., .sup.2 H and .sup.3 H, are heated to extremely high temperatures in order to form a hot plasma. The toroidal magnetic field B.sub.T confines the plasma to a toroidal volume inside of the vessel 22 that does not touch the walls of the vessel. This occurs because the toroidal magnetic field B.sub.T has lines of magnetic force coincident or parallel with the minor axis 25, and plasma, as a whole, is substantially confined to and follows magnetic lines of force, forming as it were a plasma ring. The poloidal magnetic field B.sub.P is needed to complete the plasma confinement against drifts caused by gradients in B.sub.T. The combined fields form, as it were, a plasma and magnetic vortex. The externally applied component of B.sub.P is also used to shape the cross sectional area of the plasma ring within the toroidal plasma volume to a desired shape. For example, at some points within the vessel, or at some times when the plasma is within the vessel, the cross sectional area of the plasma cloud may be "squeezed" thereby compressing the plasma into a smaller volume, and further increasing its temperature. At other points within the vessel, or at other times, the cross sectional shape of the plasma cloud may be expanded, with some of the plasma particles being diverted away from the main plasma body. Such control of the cross-sectional shape of the plasma cloud is, as indicated, controlled by the poloidal field coils 30. For this reason, such coils are sometimes referred to as the "shaping field coils" or "shaping field windings", Of particular relevance to the present invention, the poloidal magnetic field B.sub.P is also used as a secondary magnetic field to divert some of the plasma out of the main plasma cloud or body to a suitable target, where the plasma can be neutralized and removed from the vessel 22. FIG. 2 diagrammatically illustrates a cross section of the vacuum vessel 22 of the tokamak of FIG. 1, or similar tokamak, and shows representative magnetic poloidal field lines 38 and 39 that confine the plasma to the inner portions of the vacuum vessel. Most of the plasma is thus confined to an area shaped by the closed magnetic field lines 38 or 39. This portion of the cross section is thus often referred to as the plasma core. Because the plasma torus is symmetric about the major axis 24, the poloidal magnetic field lines also equivalently define toroidal surfaces, called magnetic surfaces. The circle-dot " " 28 shown in FIG. 2 is conventional vector notation to show the direction of the toroidal magnetic field. In FIG. 2, the shows the standard B.sub.T direction in the DIII-D tokamak, which is pointing out of the plane of the paper toward the viewer. The magnetic field line 40, which represents the transition between magnetic field lines that are closed and magnetic field lines that are open, is known as the "separatrix". The separatrix 40 thus defines the boundary or magnetic surface separating the plasma that is confined within the core of the tokamak 20 (or other plasma-confining structure) and the plasma that is not confined within the tokamak. As seen in FIG. 2, the separatrix 40 crosses near the bottom of the vessel 22 at a cross point 42. The cross point 42 is also referred to as the separatrix "X-point". Any plasma on the outside of the separatrix 40 is thus not confined, and will eventually be diverted, following other magnetic force lines, such as the open magnetic force lines 43 and 44, or equivalently, open magnetic surfaces, away from the main body of plasma to the inside edges of the vessel 22. In practice, the plasma that escapes across the separatrix flows along the magnetic field lines in a thin layer 46 just outside of the separatrix 40, known as the "scrape-off layer" (SOL), until it reaches the divertor targets 48 and/or 50. The divertor targets 48 or 50 are made from any suitable material, such as graphite, adapted to absorbed the heat of the diverted plasma and neutralize the plasma to form a gas. The gas is then collected in a plenum 52. Referring next to FIG. 3, a diagrammatic representation of a divertor electrode and pumping plenum made in accordance with the present invention is illustrated. The lower portion of the vessel 22, including the lower portion of the separatrix 40 and the X-point 42 are schematically depicted in FIG. 3. The floor of the vessel 22 in the vicinity of the X-point 42 and separatrix 40 is referred to as the divertor floor. A ring electrode 56, toroidally symmetric with the major axis 24 of the tokamak, is placed at the entrance of the a gas plenum 54. The ring electrode 56 may include cooling channels 58, through which a suitable coolant, such as water, may be circulated. In a preferred embodiment, used with the DIII-D tokamak, located at General Atomics, San Diego, Calif., the ring electrode 56 has a 20 kA capability, and is covered with a graphite armor. A liquid helium (He) cryogenic pump may optionally be positioned within the plenum volume 54. Details associated with the design and construction of the divertor electrode 56 as shown in FIG. 3, and as implemented in the DIII-D tokamak at General Atomics, are documented by Smith, J., "Design of the DIII-D Advanced Divertor", Proceedings 13th IEEE Symposium on Fusion Engineering (Oct. 2-6, 1989), pp. 1315-18, incorporated herein by reference. It is sufficient for purposes of the present invention to note that the electrode 56 is electrically insulated laterally from the vacuum vessel 22 by two plasma-facing rings 63 and 64 made of boron nitride (BN). Boron nitride material from which such rings can be made is commercially available from various suppliers, such as Union Carbide Corporation. The lateral ring insulator 64 is placed at the entrance of the plenum 54 opposite the ring electrode 56. Additional insulators 60 and 62 are used to insulate the lower and outer surfaces of the ring electrode 56, respectively. The insulators 60 and 62 may be made from any suitable electrically insulative material suitable for high temperature use, such as mica and Al.sub.2 O.sub.3 pieces. It is noted that in FIG. 3, and other similar figures presented herein, only one divertor is illustrated, where the term "divertor" refers to the combination of the ring electrode 56, insulators 60, 62, 63 and 64, and entrance aperature to the plenum 54, or other ducting. The diverter illustrated in the figures is the "outer" divertor, it being located farthest from the primary axis 24, and nearest the outer wall of the vessel 22. It is to be understood that another divertor, termed the inner divertor, may also be used that is closest to the primary axis 24, and nearest the inner wall of the vessel 22. A power supply 66 applies electrical power (sometimes referred to as the "bias potential") between the electrode 56 and the conductive wall of the vacuum vessel 22. The inner vacuum vessel surfaces that interact with plasma are covered with graphite tiles, all of which are in electrical contact with the vessel 22. During operation, the ring electrode 56 is not heated, and does not reach temperatures at which thermionic emission is important. The coolant channels in the ring electrode help to maintain its temperature and prevent overheating, especially when the electrode 56 is also used as a divertor plate or target, as is the case for a preferred embodiment of the invention. In such preferred embodiment, used with the DIII-D tokamak, four toroidally distributed feed conductors are also used to connect the power source 66 to the ring electrode 56 to ensure that local magnetic errors are small, even at the 20 kA design maximum electrode current. As is known in the art, the separatrix strike position is controlled by the X-point location, which in turn is controlled by the current distribution in the lower poloidal field shaping coils. Advantageously, by controlling such current distribution in an appropriate manner, the outer strike position can be varied smoothly from the middle of the divertor floor to the upper surface of the ring electrode 56. Such feature permits the separatrix strike position to be swept across the divertor floor, thereby reducing the heat flux absorbed in the divertor floor by time averaging. Unfortunately, the gas static pressure in the plenum 54 is very geometry sensitive, i.e., it varies as a function of the position of sweeping X-point. To demonstrate this dependence, the gas static pressure was examined as a function of separatrix position by sweeping the X-point 42 slowly in the radial direction (over a time period of about 1600 msec) while maintaining otherwise unvarying plasma conditions. During this study, the potential of the electrode 56 was maintained at the same potential as the vessel structure 22, so as not to influence the plasma. The results of such a study, made using the DIII-D tokamak previously referenced, are illustrated in FIG. 4. As seen in FIG. 4, the plenum gas pressure increased as the separatrix neared the entrance aperture of the plenum 54. The maximum pressure occurred when the separatrix 40 was within about 1 cm of intercepting the lower inside corner of the electrode 56 (corresponding to a time of about 2500 msec on the horizontal axis of the pressure vs. time graph included in FIG. 4). The pressure then decreased rapidly with additional outward displacement. Pressures under the X-point and above the plenum remained low at all times. It is noted that plenum pressure at the optimum (highest pressure) separatrix location depends on several parameters, including plasma density and divertor surface condition. The pressure also depends on neutral beam heating power as shown in FIG. 5. The plenum gas pressures shown in FIG. 5 are for H-mode operation and are the maximum values measured during slow divertor sweeps like the one illustrated in FIG. 4. As seen in FIG. 5, pressures on the order of 10 mtorr are typically obtained, and such pressures are sufficient for practical high-throughput pumping in a tokamak, such as the DIII-D tokamak. Another factor influencing the plenum pressure is electrode bias. The effect of applied electrode bias is depicted in FIG. 6, where plenum pressure for Ohmic, L-mode and ELMing H-mode single-null discharges using the standard DIII-D tokamak toroidal magnetic field and plasma current directions is depicted. (Note that "ELMing" refers to the presence of Edge Localized instability Modes. ELMs are a common feature of H-mode operation and are described in the "ASDEX Team" reference, cited previously.) As seen in FIG. 6, negative electrode potential relative to the vacuum vessel increases the plenum gas pressure, while positive bias potential decreases the plenum gas pressure. Further, it has been observed by other measurements not illustrated in FIG. 6 that negative electrode potential decreases particle recycling at the inner divertor and the inner wall. Similarly, it is evident that positive bias potential increases particle recycling at the inner divertor and inner wall. These effects are qualitatively present, regardless of whether the separatrix strikes the electrode 56 or the vessel floor, so long as the separatrix 40 is close enough for the diverter plasma to interact with the biased electrode 56. As was the case for the data presented in FIG. 5, gas pressure remains low under the X-point at all times. The bias electrical power used for the experiments shown in FIG. 6 was approximately 1 MW. Some of this power is radiated, while the remainder appears at the divertor electrode as heat. Advantageously, electrode operation does not contribute to plasma impurities. The data presented in FIG. 6 was measured at the time of maximum plenum gas pressure during slow divertor sweeping as in FIG. 4 at two heating powers for medium density (2.3.about.2.5.times.10.sup.19 m.sup.-3) Ohmic and L-mode plasmas, all at B.sub.T =2.1 T, I.sub.p =1 MA, and standard field directions. As seen from the data in FIG. 6, the dependence on bias potential appears to be monotonic in all cases where data is available. Positive bias decreases the plenum pressure, and negative bias increases the pressure. Reversal of the toroidal field direction reverses the roles of positive and negative bias. It is noted that the observed changes to the particle recycling and plenum pressure are qualitatively consistent with the expected consequences of bias-induced drift velocity V.sub.E =E.times.B.sub.T /B.sub.T .sup.2 in the SOL as suggested by the early experiments of Strait, cited above. Thus, for example, as shown in FIG. 7, which is drawn for standard DIII-D field directions, the bias voltage establishes a poloidal electric field E.sub.P within the magnetic surfaces contacting the electrode 56. The resulting radial V.sub.E yields the observed recycling changes at the inner wall of the vessel 22. As also seen in FIG. 7, a radial electric field E.sub.r is established normal to the biased magnetic surfaces. When the separatrix 40 strikes below the electrode 56, the resulting E.sub.r .times.B.sub.T produces a poloidal flow in the SOL between the separatrix and the biased surface, directing plasma toward the inner divertor for positive bias potential, and toward the outer divertor for negative bias potential. This is again in qualitative agreement with the recycling and plenum pressure observations. When the separatrix strikes the electrode 56, the electric potential distribution becomes more complicated. Such potential distribution is shown qualitatively in FIG. 8. A large E.sub.r appears as a consequence of potential jumps across a thin boundary layer near the separatrix 40. This is so because just inside the separatrix, E.sub.P must be small, so the potential difference between large and small major radius SOL must appear across a boundary layer. A large E.sub.r also appears in the boundary layers separating biased magnetic surfaces from surfaces contacting insulators at one end and the vessel at the other, such as the boundary layer 72. Because the leakage of current across B is small, any surfaces contacting the vessel at one end and an insulator at the other remain close to vessel potential, FIG. 8 qualitatively depicts how the E.times.B.sub.T flow, which is along equipotential surfaces, drives plasma across the SOL, along paths 72; across the divertor separatrix along paths 74; across the X-point 42; and across the "arch-shaped" surfaces below the X-point, following paths 76. Then, the large E.sub.r in the boundary layer that grazes the lower inside corner of the electrode 56 drives the plasma rapidly along path 78 into the plenum entrance. This driving of the plasma thus allows the electrode/separatrix geometry shown in FIG. 8 to function as an efficent plasma pump. It is noted that although the plasma striking the electrode 56 before reaching the plenum entrance is neutralized, most of the neutral atoms are reionized near the neutralization site by the dense divertor plasma, and the reborn ions are thereafter subject to the same E.times.B drifts. Hence, all outer SOL ions, except those that get buried in the electrode and floor material, eventually reach the plenum entrance at the optimum position for pumping. This reionization and subsequent continued flow as plasma was absent from the early experiments of Strait, referenced previously. Advantageously, it has been determined experimentally that the above described process of E.times.B drift significantly reduces the sensitivity of the plenum pressure dependence on separatrix strike position, most notably when the separatrix strikes the electrode. Data demonstrating this feature is shown in FIG. 9, which shows a data plot comparing plenum gas pressure with neutral beam power during H-mode operation for three divertor conditions. The "+" marks are data for the separatrix diverting plasma into the entrance aperture with no bias applied to the divertor electrode 56 as in FIG. 5 (separatrix not striking the electrode, optimum positioning for maximum pressure). The "O" marks are data for the separatrix striking the electrode, still with no bias applied to the electrode 56. The " " marks are data for the separatrix striking the upper inside corner of the electrode, and with bias applied to the electrode 56. As seen from the data presented in FIG. 9, even when the separatrix strikes the upper inside corner of the electrode (which for the DIII-D where the data were taken is 7 cm above the top of the plenum entrance aperture), the plenum pressure with bias is almost as high as the pressure for the separatrix optimally positioned in the aperture without bias. Further, when the separatrix is in the high position, the pressures with bias are about four times higher than those without. Thus, the biased divertor acts as an E.times.B plasma pump, particularly when the separatrix strikes the electrode. The operation of the plasma pump of the present invention may be analyzed by considering a potential difference "V" applied across a plasma gap of width "w". The pump cross section is Lw, where "L" is the dimension in the magnetic field direction. The pump speed, S.sub.E, may then be expressed as EQU S.sub.E =Lwv.sub.E =LwE/B.sub.T =LV/B.sub.T, where EQU V=.intg.E.multidot.dw.apprxeq.wE. Note that the v.sub.r and v.sub.P pumping speeds are approximately the same. For the experiments described, L=2.pi.R.apprxeq.10 m; V/B.apprxeq.(200 V)/(2T)=100 m.sup.2 /s, so S.sub.E .apprxeq.1000 m.sup.3 /s. For comparison, the mechanical vacuum pumps presently installed on the DIII-D tokamak have a combined speed of only about 10 m.sup.3 /s. It is noted that the steady state plasma density in the DIII-D tokamak during operation in the H-mode is proportional to the current, and depends on little else. The density cannot be varied appreciably by normal operational procedures. However, modest controlled density changes were observed under conditions of strong divertor bias, as shown by the data presented in FIG. 10. FIG. 10 illustrates the applied bias potential, average density, and average temperature during H-mode operation as a function of time. As seen in FIG. 10, for normal field directions, negative bias reduced plasma density and positive bias increased plasma density. The density changes were achieved, even though the planned pumps have not yet been installed in the divertor plenum 52. The density changes were accompanied by reciprocal changes in the volume-averaged temperature (average of T.sub.e and T.sub.i) such that the total energy and energy confinement time remained nearly constant. Hence, as indicated above, it is seen that the present invention offers an excellent means for particle control in a tokamak, or similar plasma-confinement apparatus. Neutral gas pressures on the order of 10 mtorr are possible during operation in the H-mode when optimum separatrix position is maintained. By using the plasma pump of the present invention, the E.times.B drift also transports particles across the SOL and optimally into the plenum entrance when the separatrix strikes the electrode (which is far removed from the optimum positioning of the separatrix). Hence, through the selective application of a bias potential to the divertor electrode, the divertor acts as a geometry-insensitive high capacity pump to drive plasma into the plenum aperture. Such plasma pump can advantageously be used to reduce vacuum pumping requirements for steady state plasmas; to exhaust plasma from low density plasmas; to establish low collisionality, low density H-mode plasmas for current drive; and to make plasma exhaust insensitive to divertor geometry, especially to the variable geometry of swept divertors. While the invention herein disclosed has been described by means of specific embodiments and applications thereof, numerous modifications and variations could be made thereto by those skilled in the art without departing from the scope of the invention set forth in the claims.
051611791	abstract	A beryllium window comprises a disk-shaped beryllium plate containing beryllium as an essential element, a welding film partially merged into the outer peripheral portion of the beryllium plate and formed of a substance having at least one element selected from the group consisting of silver, gold, nickel and copper, and a reinforcing unit of a stainless steel, and the welding film is partially merged into the reinforcing unit, wherein the welding film fixes the beryllium plate to the reinforcing unit through diffusion welding so that the beryllium window is less deformative against heat stress.
	description	Referring now to the figures of the drawings in detail and first, particularly, to FIG. 6 thereof, there is shown a reactor core, indicated by the reactor pressure vessel 64 together with the fuel elements 65, 66. The reactor core is located below a water level 60 of a reactor pool 61 which is connected to a storage pool 63 via a channel 62. A loading machine 70 is movable through the use of a carrying structure 67. This loading machine carries a hollow mast divided into two parts 71, 72 which are provided parallel to one another and which are movable relative to one another in a horizontal plane and which contain their own devices for gripping, raising and lowering. These devices are illustrated symbolically in each case as a cable assembly with lifting drum 74 and with a hook 73 as a gripper. In a first step A, the loading machine is brought over the reactor core and two fuel elements 65, 66 are simultaneously or successively gripped and raised into the parts 71, 72 of the mast. If the fuel elements are closely adjacent to one another, the two mast parts are preferably in a first end position 70a, in which they form a hollow mast open at the bottom, but virtually closed laterally. It will be assumed, here, that the fuel elements are subsequently to be inspected visually and, in this case, to be rotated. This is illustrated by position 70b: the spacing of the two mast parts together with the fuel elements 65, 66 is increased and at least one mast part is rotated into a longitudinal axis. In this position, each individual fuel element can be lowered and be inspected outside the mast. In order to transport the fuel elements into the storage pool 63, preferably the first end position 70a of the two mast parts is resumed (Step C). The loading machine can be moved into this position through the water-filled channel 62 quickly and without putting the fuel elements located in its mast at risk (Step D). When the loading machine is brought over the storage rack 75 in the storage pool 63, a relative movement of the two mast parts causes their mutual spacing to be enlarged to the extent desired for their storage (Step E). Finally, in their other end position 70c (wide position), the fuel elements are lowered again, that is to say they are removed from the mast (Step F). In principle, a step sequence A, E, D, F is also possible. In the same manner, two fuel elements may also be moved out of the storage rack into the reactor core, in which case a reduction in the relative spacing takes place between the first step (raising of the fuel elements out of the storage rack into the two mast parts) and the last step (lowering of the fuel elements into the reactor core). Since the fuel elements are often slightly bent in the reactor pressure vessel, fuel elements can often be drawn out of the reactor core or inserted into the reactor core only when they are at the same time rotated. It may happen, for example, that an irradiated fuel element which is still usable and is temporarily set down in the storage rack has to be rotated through xe2x88x9290xc2x0, +90xc2x0 or 180xc2x0 (Step B in the diagram of FIG. 6). FIG. 1 is a diagrammatic illustration of a loading machine for handling of fuel elements. This loading machine has essentially a movable bridge 1, such as is installed in reactor buildings. In the following, a pressurized water reactor will in particular be considered here. The direction of movement of the bridge 1 leads into and out of the plane of the drawing sheet. A trolley 2 is located on the bridge. This trolley 2 is movable in the geodetically horizontal direction at right angles to the bridge 1. Located on this trolley 2 is an operating platform (not illustrated) from which the loading machine can be operated. Furthermore, a guide mast 3 is mounted rotatably about its mid-axis on the trolley 2. A centering bell 4 is located in the guide mast 3. Within this centering bell 4 there is a double gripper which has an outer fuel element gripper 5 and an inner control element gripper 6. Above a frame 2.1 of the trolley 2 are located lifting mechanisms 7 which are provided on a lifting mechanism linkage 7.1 connected to the guide mast 3. FIG. 2 is a cross sectional view of the guide mast 3, together with the centering bell 4 located in the guide mast 3 and with two fuel elements 8a, 8b. The guide mast 3 is divided in the vertical direction at the points 3.1. In other words, the guide mast is composed of two halves 3a, 3b of a tube divided in the axial direction. The centering bell 4 has a rectangular cross section and is divided vertically at the points 4.1 according to the division of the guide mast 3. Each of the portions 4a, 4b of the centering bell 4 is dimensioned in such a way that a fuel element Sa, 8b can be received in it. For a stable guidance of the centering bell 4 in the guide mast 3, the guide mast has angle irons 3.2 along its longitudinal direction. Angle irons 3.2 located diagonally opposite one another carry a roller guide 3.3. Guide wheels 4.3 configured as rollers engage with sufficient slip into these roller guides 3.3. The axes of rotation of these guide wheels 4.3 are perpendicular to the surface of the angle irons 3.2 which carries the roller guide 3.3. The guide rollers 4.3 are mounted on a virtually trapezoidal appendage 4.2 of the centering bell 4, the appendage largely overcoming the spatial distance between the centering bell wall and the angle iron 3.2. Located opposite these virtually trapezoidal appendages 4.2 are further trapezoidal appendages 4.4 of the centering bell 4. On their side adjacent to the angle iron 3.2, these appendages 4.4 have running rollers 4.5, the axis of rotation of which is oriented parallel to the adjacent surface of the angle iron 3.2. A running roller 4.6, which rolls directly on the inner wall of the guide mast 3, is in each case mounted on the outer wall of the centering bell 4, on the two remaining free outer sides of the centering bell 4. The fuel elements 8a, 8b or 8 (FIG. 3) are received in each part 4a, 4b of the centering bell 4 through the use of fuel element grippers 5 not illustrated in FIG. 2. In this case, each of the two centering bell halves is provided with a fuel element gripper 5 (FIG. 3). One of these fuel element grippers 5 is supplemented by a control rod gripper 6 provided inside the fuel element gripper 5, to form a so-called double gripper according to Published German Patent Application DE 17 64 176. Reference is made to DE 17 64 176 with regard to details of the double gripper. The double gripper has essentially two functional elements, specifically a fuel element gripper 5 and a control element gripper 6 provided concentrically in the fuel element gripper 5. The basic configuration of the double gripper is explained below in context with the function of the double gripper: In order to extract a fuel element 8 from the reactor core or a fuel element storage rack, or else extract a control element 9 from a fuel element 8, first the entire gripping configuration is brought into a position above the respective fuel element with the aid of the bridge 1 and the trolley 2. The centering bell 4 is then lowered until its lower edge assumes a position just above the fuel element 8. When a fuel element 8 is extracted from a reactor core, the centering bolts 4.7 of the centering bell 4 engage into corresponding bores in the fuel element heads of the fuel elements adjacent to the respective fuel element 8. Exact positioning of the fuel element gripper 5 and of the control element gripper 6 is brought about in this way. At the same time, the centering bell is held in its position in the mast in such a way that its weight does not act on the fuel elements which are located geodetically below it. The fuel element gripper 5 is subsequently lowered until it latches with its gripping latches 5.1 into the fuel element head of a fuel element 8. The fuel element 8 held by the fuel element gripper 5 is then lifted upward and moved into the centering bell 4 with the aid of a lifting mechanism 7 and a double cable 7.2. A control element can also be extracted from a fuel element 8 in a similar way. For this purpose, after the centering bell 4 has been moved down, the fuel element gripper 5 is interlocked in an upper position in the centering bell, so that it cannot move down to the fuel element 5. The control element gripper 6 is subsequently lowered until, with a control element gripper head, it grips the head of a control element contained in the fuel element 8. The control element gripper 6 catches with the control element by using the gripper latches contained in the fuel element gripper head. The control element is then drawn out of the fuel element 8 into the centering bell 4 by the raising of the control element gripper 6 via a lifting mechanism 7 and the double cable 7.2. After either a fuel element 8 or a control element has been received in the centering bell 4 in this way, the centering bell 4 itself is drawn upward and moved into the guide mast. The fuel element 8 or the control element can then be moved horizontally in the reactor space. The other half of the centering bell 4 is equipped with a simple fuel element gripper in the manner of that described. For the transport of fuel elements and control elements in the reactor space, it is usually sufficient for only one of the two grippers in the centering bell 4 to be configured as a double gripper. It is, of course, also possible for both grippers in the centering bell 4 to be configured as double grippers. Since fuel elements are positioned very close to one another in the reactor core, but, on the other hand, a given space remains between the individual fuel elements in the fuel element storage rack, it is necessary to have the possibility of varying relative to one another the position of the two fuel elements transported through the use of the device described. The same need arises when a visual check of the fuel elements is to be carried out. In order to allow a change in the position of the fuel elements 8a, 8b relative to one another, the guide mast 3 is divided vertically at the points 3.1 and the centering bell 4 is likewise divided vertically at the points 4.1. As illustrated in FIGS. 4a-4c, each half 3a, 3b of the guide mast 3 and the corresponding half 4a, 4b of the centering bell 4 are equipped with a lifting mechanism 7a, 7b. The guide mast halves 3a, 3b, together with their internal fittings, such as the centering bells, halves 4a, 4b and grippers, and their associated lifting mechanisms 7a, 7b, are movable horizontally on the trolley 2. It is also sufficient, however, if only one of the two guide mast halves is movable on the trolley 2 and the other half is fixed. The movement device for the two guide mast halves is indicated in FIG. 1 by a guide 2xe2x80x3 which is movable in the x-y direction and in which wheels 2xe2x80x2 for rotating the structure 2.1 are guided. Rotation of the fuel elements 8b about their own axis thereby becomes possible at the same time. For this purpose, the guide mast halves are moved through the use of the guide 2xe2x80x3 until the mid-axis of the transported fuel element is congruent with the axis of rotation. When the rotary shield then executes a rotational movement, the fuel element, together with the associated guide mast half, the centering bell half and the gripper, is corotated. The possible angles of rotation amount, in this case, to +90xc2x0, xe2x88x9290xc2x0 and 180xc2x0. In order to avoid that the structures or installations on the guide mast, in particular the lifting mechanisms, do not collide during the rotational movement, it may be necessary that the horizontal movement exceeds that distance which is required to provide sufficient spacing between the two transported fuel elements for placing these fuel elements into a fuel element storage rack. In order to increase the handleability of the loading machine, the transported fuel elements can be moved toward one another again after rotation has taken place. The operation just described is illustrated diagrammatically in FIGS. 4a-4c and 5a-5c. Thus, FIGS. 4a-4c show the first lifting mechanism 7a and the second lifting mechanism 7b, together with the two transported fuel elements 8a and 8b which are moved apart (and, with them, also the guide mast halves 3a, 3b and centering bell halves 4a, 4b). At the same time, the fuel element 8b has been rotated through 180xc2x0 along its mid-axis in the rotary position A. The fuel element 8b, together with its lifting mechanism 7b and the associated guide mast half and also with the centering bell half, have subsequently been moved back toward the other fuel element 8a into position B indicated in FIG. 4. A similar procedure was performed in FIGS. 5a-5c. Here, after the fuel elements 8a and 8b, together with the associated guide mast half 3 and centering bell half 4, were moved apart, the fuel element 8b was rotated through 90xc2x0 about the longitudinal axis of the fuel element 8b (Position Axe2x80x2) and thereafter moved back again in the direction of the fuel element 8a (Position Bxe2x80x2). In both cases, the fuel elements 8a, 8b, which were in the narrow position in the core (FIG. 2), are oriented again, with sides parallel to one another, in the wide position (Position B in FIGS. 4a-4c, Position Bxe2x80x2 in FIGS. 5a-5c), the spacing of the fuel elements corresponding to the spacing which they must have, for example, in the fuel element storage rack, on a test stand for inspection, maintenance or repair or when they are to be reinserted into the reactor pressure vessel. However, their orientation relative to one another is different. The situation may arise where a fuel element has been bent during the operation of the reactor and cannot readily be removed from the formation it forms with the adjacent fuel elements or inserted into the formation. However, removing and inserting becomes possible when the fuel element is rotated through +90xc2x0, xe2x88x9290xc2x0 or 180xc2x0. Such faults in which the vertical movement of the fuel element is impeded by obstacles can be detected if the lifting mechanism is mounted on a platform having a weight measuring device which signals both an improper decrease in weight (the fuel element sits on an obstacle) and an increase in weight (the fuel element is detained) during the vertical movement. Moreover, it may be advantageous if the fuel element can be rotated at the workplace so that it can be inspected, attended to or worked on from different sides. For an inspection, the trolley 2, together with the guide mast halves fastened to it and with the elements held therein, is moved over the new workplace and the transported fuel elements or transported control elements can be lowered out of the guide mast 3. For this purpose, the centering bell halves are detained in the moved-up position via fastening devices in the guide mast, while the fuel element gripper 5 moves downward and thus frees the fuel element 8 for inspection. A visual inspection of the transported control element may also be carried out in a corresponding way, in that the centering bell 4 and the fuel element gripper 5 are held in a moved-up position via locking devices, while the control element gripper moves downward and thus frees the control element 9. These operations are illustrated in detail in DE 17 64 176. In the loading machine described, the fuel element 8 or control element 9 can be moved out of the centering bell 4 and the guide mast 3 in that the fuel elements 8 and control rods 9 can be individually moved out downward, even when the loading machine is equipped with two fuel elements 8 or with one fuel element 8 and one control element 9. Even in positions in which they are moved apart and which correspond to FIGS. 4a-4c and 5a-5c, fuel elements 8 or control elements 9 can be moved independently of one another out of the centering bell half held high in the guide mast half and can be examined. Through the use of the loading machine illustrated in the exemplary embodiment, two fuel elements or one fuel element and one control element can be handled simultaneously. It is also possible, however, to configure the loading machine in such a way that a plurality of fuel elements, for example four fuel elements, can be handled simultaneously. For this purpose, it is necessary to divide the guide mast and centering bell into four in a similar way to the exemplary embodiment. The handled fuel elements then result in a bundle of two times two fuel elements. It is then necessary, correspondingly, that, for a mutual relative movement, the fuel elements be either movable horizontally in the horizontal direction or be movable at an angle of 45xc2x0 with respect to the dividing planes of the guide mast 3. In all cases, it is possible that the loading machine handles a number of fuel elements or control elements, which is below the capacity of the loading machine.
	claims	1. Method for loading radioactive elements in a package, comprising the successive following steps in this order:(a) placing a plurality of radioactive elements in a storage basket in a pool containing water, wherein the storage basket comprises at least one element chosen among a shielding sidewall, a sheath or a plate, for radiological protection;(b) after said step (a), extracting the basket containing the radioactive elements out of the pool, wherein the basket is configured such that a majority of the water in the basket is removed as the basket is extracted out of the pool; and(c) after said step (b), loading said basket containing the radioactive elements in the package. 2. Method according to claim 1, wherein each of the radioactive elements is held in place by gravity in the storage basket. 3. Method according to claim 1, wherein steps (a) to (c) are repeated several times such that several baskets are loaded in said package. 4. Method according to claim 3, wherein the baskets housed in the package together define an external lateral surface that is approximately complementary to the lateral surface of a cavity of the package in which the baskets are housed. 5. Method according to claim 1, wherein the package comprises at least one element chosen among a solid lower part or two concentric shells, for radiological protection. 6. Method according to claim 1, wherein the package is closed by a lid after said basket containing the radioactive elements has been loaded in the package cavity. 7. Method according to claim 1, wherein each storage basket houses between five and ten radioactive elements. 8. Method according to claim 1, wherein said radioactive elements are worn rod cluster control guides or irradiated fuel assemblies. 9. Method of transport of radioactive elements including successive following steps in this order:(a) placing a plurality of radioactive elements in a storage basket provided with radiological protection means, in a pool containing water;(b) after said step (a), extracting the basket containing the radioactive elements out of the pool, wherein said basket is configured such that a majority of the water in the basket is removed as the basket is extracted from the pool;(c) after said step (b), loading said basket containing the radioactive elements in a package; and(d) after said step (c), transporting the package comprising the basket. 10. Method of interim storage of radioactive elements including the successive following steps in this order:(a) placing a plurality of radioactive elements in a storage basket provided with radiological protection means, in a pool containing water;(b) after said step (a), extracting the basket containing the radioactive elements out of the pool, wherein said basket is configured such that a majority of the water in the basket is removed as the basket is extracted from the pool;(c) after said step (b), loading said basket containing the radioactive elements in a package; and(d) after said step (c), stocking the package comprising the basket during interim storage. 11. Method according to claim 1, wherein the storage basket is made of a material that provides radiological protection to an area surrounding the storage basket. 12. Method according to claim 1, wherein said step (c) of loading said storage basket containing the radioactive elements in the package occurs outside of the pool.
	summary
041773853	summary	BACKGROUND OF THE INVENTION This invention relates to nuclear fuel storage and in particular to a method and apparatus for storing fuel assemblies in a pool. Reactor fuel element assemblies are frequently stored in storage pools which can accommodate either new or spent fuel assemblies. The pool is filled with water which may be borated. This supplies cooling of the assemblies as well as moderator and also poison if the water is borated. It is of course essential that the stored mass not be permitted to assume a geometry which is either critical or supercritical. The storage pool must be provided during the initial construction of the plant so as to provide for storage of any fuel assemblies which would have to be removed from the reactor. The storage pool at this time need not be capable of storing its ultimate capacity. Investment in expensive materials as components of the storage assembly require a present investment if supplied with the initial storage rack. There is an obvious economic saving if such investment could be deferred. Most storage arrangements are designed for a particular fuel enrichment and are, therefore, completely inadequate should fuel of additional enrichment have to be stored at some time in the future. While borated water may be used in the pool to compensate for this additional enrichment it is considered an unsafe practice to completely rely on the boron content. In the event that the pool develops a leak and water must be replaced with fresh water, the boron content is depleted. Furthermore, there is always a potential for an operating errorwhereby the boron concentration is not maintained at the safe level. Storage racks have been designed utilizing the flux trap principle as illustrated in U.S. Pat. No. 4,004,154 issued to Frank Bevilacqua on Jan. 18, 1977. In such a device a stainless steel plate closely surrounds the fuel assembly being stored with water contained between the plates. Fast neutrons from the fuel pass through the plates and are slowed to thermal levels by the water. At the thermal level they are not able to return through the plates to the fuel. The required spacing for a particular fuel enrichment is calculated according to well known nuclear physics principles. There is an inherent expense in holding tolerances of a structure where multiple plates are involved and tolerances must be simultaneously held. SUMMARY OF THE INVENTION It is an object of the invention to defer a portion of the investment in a fuel storage rack for some years until increased capacity of the rack is required. It is further an object to obtain the ability to store fuel of an enrichment greater than that which has been forecast. It is a further object to reduce the cost of a flux trap type storage rack. A nuclear fuel storage apparatus for use in a water filled pool is fabricated of a material such as stainless steel in a form of an egg crate structure having vertically extending openings. Adjacent openings have a common wall between them which extends throughout the height of the active length of the fuel to be stored. Fuel may be stored in this basic structure in a checkerboard pattern with high enrichment fuel or in all openings when the fuel is of low effective enrichment. A plurality of inserts of a material such as stainless steel are adapted to fit within these openings. The inserts have two plates, one parallel to each of two adjacent sides of an opening, and the plates extend throughout a length generally equal to or greater than the active length of the fuel to be stored. The plates are stored in a similar location in each opening so that a water gap and, therefore, a flux trap is formed between adjacent fuel storage locations. These inserts may be added at a later time and fuel of a higher enrichment may be stored in each opening. When it is desired to store fuel of still greater enrichment, poison plates may be added to the water gap formed by the installed insert plates, or substituted for the insert plates. Alternately or in addition thereto fuel may be installed in high neutron absorption poison boxes which surround the fuel assembly to store fuel or still greater enrichment. It is normally expected that the inserts must be removed at this time because of physical problems of storing the same size fuel assembly with a surrounding box. Stainless steel boxes installed in this manner would function to produce an effective flux trap. The stainless steel inserts and the poison plates are each not required until the capacity of the basic egg crate structure is approached. Purchase of these items can, therefore, be deferred for many years. Should the fuel to be stored be of higher enrichment than initially forecast, the deferred decision on the poison plates makes it possible to obtain increased poison in the plates to satisfy the newly discovered requirement. Even if the storage rack were to be initially supplied with all the inserts in place in accordance with the flux trap principle, construction costs may be reduced. While basic tolerances must still be maintained on the original egg crate structure, the inserts are formed with their own tolerances, which while related to the basic structure tolerances need not be simultaneously maintained.
	description	This Application claims priority to U.S. Provisional Patent Application No. 62/384,490 filed on Sep. 7, 2016, which is hereby incorporated by reference in its entirety. Priority is claimed pursuant to 35 U.S.C. § 119 and any other applicable statute. This invention was made with Government support under Grant Nos. 2R44MH097271, R21AG049918, and HHSN261201400041C, awarded by the National Institutes of Health. The Government has certain rights in the invention. The technical field generally relates to devices and methods used in the automated preparation of radiopharmaceuticals including Positron Emission Tomography (PET) probes. The advent of molecular imaging approaches such as Positron Emission Tomography (PET) has enabled measurements of molecular and cellular mechanisms throughout the body in preclinical and clinical settings. Such measurements have widespread diagnostic utility and their use for evaluation of treatment responses and to assist drug development is expanding rapidly. Probes are traditionally synthesized by skilled radiochemists using specialized equipment and facilities that reduce their radiation exposure when working with large quantities of short-lived isotopes necessary to produce a final dose sufficient for imaging a human. In recent years, the development of automated radiosynthesizers that can produce a variety of different probes with minimal human intervention or radiation exposure has aimed to simplify routine synthesis of PET probes, especially for the clinic. As such, these synthesizers can be operated by technicians and do not require a highly trained radiochemist. Additionally, some automated systems can be configured to prepare different PET probes and thus also act as valuable tools for researchers developing new synthesis protocols for novel probes. For example, the ELIXYS radiosynthesizer (Sofie Biosciences, Inc., Culver City, Calif.) is a disposable cassette-based, automated multi-reactor radiosynthesizer that is designed for both the development of new synthesis protocols as well as routine clinical and pre-clinical probe production. While synthesis operations for PET probes have been automated, once the probe has been produced, the final product that is injected into the subject often requires subsequent purification and formulation to remove or reduce exposure to potentially toxic organic solvents and chemical impurities. In some synthesis operations, the output of automated synthesizers is coupled to an entirely different purification system (e.g., high performance liquid chromatography HPLC) that is run by its own separate automated control system. After purification, formulation and concentration of the PET probe is performed manually using, for example, bulky rotary evaporation equipment. FIG. 1 illustrates a sequence of operations used to generate an injectable PET tracer according to the prior art. Thus, users had to employ multiple different types of systems to produce a final, injectable product. Not only is this expensive but it also means that users have to switch between different control systems for the various sub-systems, and the equipment takes up valuable space within the lead-shielded hot cell where the radiochemistry takes place. Different computers and control software are needed for each process making the overall automation process more complicated and expensive. In one embodiment, a device for purifying and formulating a radiopharmaceutical compound such as a PET tracer including an automated purification subsystem and an automated formulation subsystem. These two subsystems are contained or housed within a single device and are controlled using a computer controller that interfaces with the device. In one embodiment, the controller is the same controller that is used to control operations of the radiosynthesizer. The purification subsystem is used to take a crude radioactive product that has been generated from a radiosynthesizer and load the same into a sample loop (ins some embodiments one of a plurality of loops) using an automated HPLC injection valve. The crude product in the sample loop is delivered to one of a plurality of columns after passing through a column selector valve. After separation in the column, the product components (e.g., product, contaminants, residual reactants) are detected using an in-line UV detector and radiation detector. A computer controlled downstream fraction collection valve is actuated to pass these fractional components to waste, one or more fraction collection containers (e.g., tubes or vials), or a product output line. The automated formulation subsystem includes a dilution reservoir that receives the fraction or product contained in the product output line (i.e., the product that is to be formulated). The diluted fraction or product is pushed onto a solid-phase extraction (SPE) cartridge using a compressed source of inert gas that enters the dilution reservoir. The fraction or product becomes trapped on the sorbent material (e.g., resin) contained therein while the liquid in which the product is dissolved passes through the resin and into a waste container. A multi-port syringe pump is provided that includes an output line that connects to a computer controlled cartridge valve that is located upstream of the SPE cartridge. The pump is used to first rinse the SPE with water to remove impurities or organic solvents. Next, using a different input port of the pump, an eluting liquid such as ethanol is aspirated and then pumped through the SPE cartridge to release the trapped product. In one embodiment, the eluted fluid that contains the fraction or product of interest is transferred to a final product container (e.g., a vial). Next, the pump then pumps a saline or other aqueous solution through the SPE cartridge and into the final product container for reformation (e.g., to ensure that ethanol content is below the allowable limit). The radiopharmaceutical compound contained in the final product container is ready for use. According to one embodiment of the invention, a device for purifying and formulating a radiopharmaceutical compound such as a PET tracer includes an automated purification subsystem that includes a computer controlled injection valve coupled to a high performance liquid chromatography (HPLC) pump, a plurality of sample loops, and an output line from the injection valve, the computer controlled injection valve having one or more ports configured to receive an input fluid containing the radiopharmaceutical compound, wherein one of the plurality of sample loops is connected to the HPLC pump and the output line from the injection valve and another sample loop is connected to the port configured to receive the input fluid. An automated column selector valve is coupled to the output line from the injection valve and is configured to select one of a plurality of columns for fluid to pass through. The output from the column goes first through a UV detector and a radiation detector to a downstream fraction selector valve that is located downstream of the detectors and configured to divert fluid flow to one of a product output, waste output, and fraction output. The device also includes an automated formulation subsystem coupled to an output of the downstream fraction selector valve that contains the desired product to be formulated. The automated formulation subsystem includes a dilution reservoir configured to receive the product fraction from the downstream fraction selector valve, the dilution reservoir being fluidically coupled to a solid-phase extraction cartridge. A computer controlled pump (e.g., syringe pump) is coupled to a plurality of different fluid reagents that include a wash solution, a saline solution, and an eluting solution, the computer controlled pump configured to pump selected the fluid reagents through the solid-phase extraction cartridge via a computer controlled cartridge valve interposed between the dilution reservoir and the solid-phase extraction cartridge. A final output line is fluidically coupled to the output of the solid-phase extraction cartridge, wherein a computer controlled waste valve is located downstream of the solid-phase extraction cartridge and can select between directing the fluid path to waste or to a final product container. Product is first trapped in the SPE cartridge which is then followed by rinsing the SPE cartridge with water. The trapped product is then eluted off of the SPE cartridge using an eluting liquid. This eluted product may be transferred to a final product container which can then be reformulated by passing a saline or other aqueous solution into the final product container. Alternatively, the eluted product may be transferred back to the automated radiosynthesizer for subsequent chemical reactions. In the latter configuration, the device for purifying and formulating a radiopharmaceutical compound is used as an intermediate step in the chemical synthesis. In another embodiment, a system for the formation, purification, and formulation of a radiopharmaceutical compound is disclosed that includes a radiosynthesizer device configured for synthesizing radiopharmaceutical compound, an automated purification subsystem, an automated formulation subsystem, and a computer accessible controller interfacing the with the radiosynthesizer device, the automated purification subsystem, and the automated formulation subsystem, wherein one or more operations of the automated purification subsystem, the automated formulation subsystem, and the controller are programmable by a user. The automated purification subsystem includes a computer controlled injection valve coupled to a high performance liquid chromatography (HPLC) pump, one or more sample loops, and an output line from the injection valve, the computer controlled injection valve having one or more ports configured to receive an input fluid containing the radiopharmaceutical compound from the radiosynthesizer device, wherein one of the sample loops is connected to the HPLC pump and the output line. An automated column selector valve is coupled to the output line from the injection valve and configured to select one of a plurality of columns to be used for sample purification. The column output is then directed into one or more detectors (e.g., a UV detector and radiation detector) for sample detection. A downstream fraction selector valve is configured to divert fluid flow to one of a product output, waste output, and fraction output. The fractions that are collected can be used by the chemist or operator to determine, for example, elution times for various products. The fractions may also be used to analyze product purity. Fractions can also be analyzed for new probe development. Fraction analysis is also used to tailor or optimize the conditions for separation of the desired products contained in the sample. The automated formulation subsystem is coupled to the product output of the downstream fraction selector valve and includes a dilution reservoir configured to receive a product fraction from the downstream fraction selector valve, the dilution reservoir fluidically coupled to a solid-phase extraction cartridge. A compressed source of inert gas is coupled to the dilution reservoir via an automated valve, wherein the compressed source of inert gas pushes fluid contents contained in the dilution reservoir into the solid-phase extraction cartridge in response to actuation of the automated valve. A computer controlled pump (e.g., syringe pump) is coupled to a plurality of different fluid reagents and configured to pump selected fluid reagents through the solid-phase extraction cartridge via a computer controlled cartridge valve interposed between the dilution reservoir and the solid-phase extraction cartridge. A final output line is fluidically coupled to an output of the solid-phase extraction cartridge, wherein a computer controlled waste valve is coupled to the final output line to divert fluid flow to waste or the final output line. A computer accessible controller interfaces the with the radiosynthesizer device, the automated purification subsystem, and the automated formulation subsystem, wherein one or more operations of the automated purification subsystem, the automated formulation subsystem, and the controller are programmable by a user. Product is trapped in the SPE cartridge followed by rinsing with water. The trapped product is then eluted off of the SPE cartridge using an eluting liquid. This eluted product may be transferred to a final product container which can then be reformulated by passing a saline or other aqueous solution into the final product container. The final product may also be transferred back into the radiosynthesizer instrument for additional chemical synthesis steps (e.g., intermediate purification). FIG. 2 illustrates an overview of a process for the formulation and concentration of a radiopharmaceutical compound (e.g., PET tracer) according to one embodiment of the invention. As seen in FIG. 2 a radioisotope is first generated or produced as seen in operation 10. The radioisotope is typically generated in a nuclear reactor, cyclotron, or generator. These radioisotopes may be produced on-site or ordered from a third party vendor. There are hundreds of radioisotopes that are used for medical applications. Of particular interest for medical imaging applications is the radioisotope fluorine-18 (F-18), which has become a key radioisotope that is used for cancer diagnosis, treatment evaluation, as well as a tool for research into cancer biology and drug development. Carbon-11 is another example of a radioisotope used in medical imaging applications (C-11). It should be understood, however, that the invention describe herein may be used with any number of radioisotopes. Next, as seen in operation 12, a radiopharmaceutical compound such as a PET tracer is synthesized using a radiosynthesizer 100 (illustrated in FIG. 3). The radiosynthesizer 100 is typically an automated device that is used to perform the chemical synthesis operations that are needed to generate the desired radiopharmaceutical compound. The radiosynthesizer 100 contains the reagents needed to generate the radiopharmaceutical compound as well as the flow paths, chemical reaction and other sites needed during the synthesis process. This includes, for example, modules or specific fluid pathways or sites for dispensing reagents, drying or evaporating products, transferring reagents or products, mixing reagents or products, reacting reagents or intermediate products, trapping, eluting, and the like. One example, of such a radiosynthesizer 100 is the ELIXYS radiosynthesizer 100 available from Sofie Biosciences (Culver City, Calif.) which is multi-reactor radiosynthesizer 100 that provides the user to perform one, two, or three pot synthesis. The ELIXYS radiosynthesizer 100 utilizes a reagent delivery robot for liquid handling and disposable cassettes 102 (FIG. 3) that provide a housing for all wetted flow paths. The ELIXYS radiosynthesizer 100 utilizes an intuitive graphical user interface which enables drag-and-drop unit operations to be performed according to the synthesis protocol being used. The ELIXYS radiosynthesizer 100 is described, for example, in Claggett et al., Simplified programming and control of automated radiosynthesizers through unit operations, EJNMMI Research, 3: 53 (2013) and in U.S. Patent Application Publication No. 2016/0280734, which are incorporated by reference herein. As seen in FIG. 2, the radiosynthesizer 100 is typically its own separate device (i.e., Device #1). While the radiosynthesizer 100 and the purification and formulation device 200 are illustrated as being two separate devices or modules that coordinate and work together, in another alternative embodiment, the functionality of the radiosynthesizer 100 and the purification and formulation device 200 may be incorporated into a single device. Still referring to FIG. 2, according to the invention, a separate purification and formulation device 200 is provided that includes functionality to purify and formulate the radiopharmaceutical compound that is generated from radiosynthesizer 100. Typically, the radiosythesizer 100 generates a final product that contains contaminants, unreacted products, undesirable solvents and the like that need to be removed prior to use in mammals (e.g., humans). For example, radiopharmaceutical compounds are often produced in organic solvents such as acetonitrile, which is toxic above certain concentrations. Prior to being used in the human body, acetonitrile must be brought down to acceptable levels. Similar toxicity profiles exist for other organic solvents and reagents used in the production of radiopharmaceutical compounds. Purification is also needed to remove unwanted radioactive compounds that could interfere with the imaging process. With reference to FIG. 2, according to the invention, the purification and formulation device 200 is provided that includes both a purification subsystem 202 as well as a formulation subsystem 204 that purifies the radiopharmaceutical compound to remove unwanted chemical byproducts or reactants as well as, in some embodiments, formulate a final solution that that contains the radiopharmaceutical compound that is ready for use in a mammal. For example, formulation may include reducing the concentration of ethanol (EtOH) to allowable levels through the dilution with saline or other aqueous solution. It may also include the addition of compounds that are used to adjust and/or stabilize pH, increase solubility, protect against radiolysis, etc. As seen in FIG. 2, this purification and formulation device 200 (i.e., Device #2) performs purification of the radiopharmaceutical compound using HPLC as seen in operation 14 followed by formulation and concentration of the radiopharmaceutical compound as illustrated in operation 16. As seen in FIG. 2, according to the invention, controller 206 is provided for the control of both the radiosynthesizer 100 and the purification and formulation device 200. As explained below, using a single computing device 208 that interfaces with the controller 206, a user is able to control and program not only the synthesis operations of the radiosynthesizer 100 but also the operations of the purification and formulation subsystems 202, 204. FIG. 3 illustrates the integration of the radiosynthesizer 100 and the purification and formulation device 200 through the use of a single controller 206. As seen in FIG. 3, both the radiosynthesizer 100 and the purification and formulation device 200 are located inside a hot cell 209. A hot cell 209 is a radiation shielded enclosure or working area that contains components that are in contact with radioactive materials. The controller 206 is located outside of the hot cell 209. The controller 206 includes a housing 212 that contains an embedded computer 214 that drives or operates the various subsystems of both the radiosynthesizer 100 and the purification and formulation device 200 using software 221. This includes, for example, driving the linear actuators, pneumatics, cooling, heating, HPLC injection, cameras, and radioactivity detectors of the radiosynthesizer 100. In addition, this includes, as explained herein, controlling the valves, pneumatics, sensors, pumps, and camera that are used in the purification and formulation device 200. The controller 206 interfaces with the radiosynthesizer 100 and the purification and formulation device 200 via data cables 216. These may include Ethernet cables and video cables when cameras are integrated into the radiosynthesizer 100 and the purification and formulation device 200. FIG. 3 further illustrates an HPLC pump 218 that interfaces fluidically with the purification and formulation device 200. The HPLC pump 218, as explained below, is used to push the generated radiopharmaceutical compound from the radiosynthesizer 100 through the purification and formulation device 200 for purification using various liquid mobile phases (e.g., up to four (4) mobile phases may be used in many HPLC pumps). Control of the HPLC pump 218 is also controlled by the controller 206 via its interface with the purification and formulation device 200. The HPLC pump 218 may be located external to the hot cell 209 (FIG. 3), internal to the hot cell 209, or located within the purification and formulation device 200. FIG. 3 further illustrates a computing device 208 that is used by the operator to interface with the controller 206. The computing device 208 may include a personal computer, laptop, tablet pc, mobile phone, or the like. The computing device 208 contains software 220 located thereon that is used by the operator to access software 221 that is run from the embedded computer 214 to select the operations to be performed by the radiosynthesizer 100 and/or the purification and formulation device 200. Typically, software 220 that includes a graphical user interface (GUI) 222 is provided on the computing device 208 so that the user can easily program the unit operations that are to be performed by the radiosynthesizer 100 and/or the purification and formulation device 200. Unit operations refer to those fundamental or building block operations that are employed the radiochemical synthesis process. Examples of unit operations include: ADD (for adding a reagent to a reaction vessel); EVAPORATE (for evaporating the contents of a reaction vessel); TRANSFER (for transferring the contents of one reactor to a next reactor; for transfer to an HPLC loop; or for transfer to the HPLC loop on the purification and formulation device 200); REACT (seals the reactor vessel to underside of disposable cartridge and heats); PROMPT (pauses sequence run and prompts the user); MOVE (moves a reactor to the front position for reaction vessel removal and/or installation and prompts the user); TRAPF18 (e.g., traps [18F]Fluoride on a quaternary methylammonium (QMA) cartridge); ELUTEF18 (uses a reagent to elute [18F]Fluoride off a QMA cartridge); MIX (mixes the contents of a reactor by stirring); EXTERNALADD (allows the user to externally add a reagent via tubing); PURIFICATION (purification of one or more columns in the purification and formulation device 200); FORMULATION (which includes four steps of (1) FORMULATION: TRAP to trap the diluted radiopharmaceutical compound on the attached SPE cartridge 252, (2) FORMULATION: RINSE to rinse the trapped compound, (3) FORMULATION: ELUTE to elute the trapped compound, and (4) FORMULATION: RECONSTITUTE which adds saline and other additives, if appropriate, to the final product to reconstitute for administration). As seen in FIG. 3, the computing device 208 may connect wirelessly (or wired) to the controller 206 using a wireless router or other connection (not shown) typically used to connect electronic devices in a wireless manner. For example, computing device 208 may connect to the controller 206 using a secure wireless Wi-Fi network or Bluetooth® connection. The software 221 that is executed on the controller 206 that is accessed by the computing device 208 may, in some embodiments, be an application or “app” 220 that is executed on the device 208. These types of application are common on tablet computers and mobile phone devices. The unit operations and their sequence are programmed by the user (e.g., by sequencing serial operations or selecting from pre-set operations) which are then executed by the controller 206 and electronics board 310 to control various computer controlled valves, pumps, and other components including, but not limited to, liquid sensors 274, 276, injection valve 240, column selector valve 242, UV detector 300, radiation detector 302 and amplifier 304, cartridge valve 316, waste valve 318, cleaning valve 320, pressure release valve 324, pressure regulator 400 set-points, syringe pump 306, and camera 314. FIG. 4 illustrates a perspective view of the purification and formulation device 200. FIG. 5 illustrates a front facing view of the purification and formulation device 200. The purification and formulation device 200 is contained within a housing 230 that can be positioned in the hot cell 209 adjacent to radiosynthesizer 100. The output line of the radiosynthesizer 100 which is typically polymer tubing is connected to the injection valve 240 of the purification and formulation device 200 via input lines coupled to cassettes 102. For example, FIG. 8 illustrates two such input lines 340, 342 which may also referred to as “output” lines from the radiosynthesizer 100. As used herein, “line” or “lines” refers to a conduit such as tubing that is used to carry fluid from one point to another. Lines can be metallic (e.g., stainless steel) as well as polymer tubing (e.g., ⅛ inch or 1/16 inch O.D. Teflon tubing). The injection valve 240, as explained in more detail herein, contains ten (10) ports with one of the ports being connected to the output line of the radiosynthesizer 100. As best seen in FIG. 9, two of the ports are connected to a first sample loop 344 while another two ports are connected to a second sample loop 346. Two additional ports on the injection valve 240 lead to waste lines 352. Another port is coupled to the HPLC pump 218. The remaining two ports on the injection valve 240 are connected to the outputs from the radiosynthesizer 100 (which serve as input lines 340, 342 to the purification and formulation device 200). For example, outputs from two different cassettes 102 are connected to the injection valve 240. The injection valve 240 is a bi-state valve and is used to change the position of flow paths. As explained herein, actuation of the injection valve 240 toggles the configuration of the sample loops 344, 346 between an “injection” position and a “load” position. Referring back to FIG. 4, the column selector valve 242 is used to select, during any particular purification run, one of a plurality of columns 243 (FIG. 8) that are held in a column holder 244 is affixed housing 230. The column holder 244 contains clips 246 that are used to hold the individual columns 243 in place. Stainless steel or another type of tubing (e.g., PEEK) is used to connect the output of the injection valve 240 to the column selector valve 242. Likewise, stainless steel or another type of tubing connects the ports on the column selector valve 242 to the individual columns 243. The column selector valve 242 includes pre-column valve components as well as post-column valve components that actuate pre and post column flow paths to selectively place the desired column 243 in the flow path. Also illustrated in FIG. 4 is a final product output fitting 248 that is used to connect to flexible tubing that either leads to a final product container 250 (as seen in FIGS. 8, 10-13) in one embodiment or, in another embodiment, leads back to the radiosynthesizer 100. This latter configuration is for when an intermediate product may need to be purified and reacted further (e.g., mid-synthesis purification). The intermediate product may undergo purification and then sent back to the radiosynthesizer 100 for additional unit operations (e.g., reactions). The fittings described herein are typically flangeless nut fittings or flangeless ferrule fittings which are commonly used for pressurized fluid applications. A SPE cartridge 252 is illustrated secured to the face of the formulation device 200 using a clip 254. Fluid flows through the SPE cartridge 252 via fittings 256, 258. FIG. 4 also illustrates the dilution reservoir 260 along with corresponding dilution reservoir fittings 262, 264, 266. Each of these fittings serves a different function. One fitting 262 is used to deliver a product fraction to the dilution reservoir 260 (i.e., product in). Another fitting 264 is used for air vent and also to deliver compressed gas (e.g., nitrogen) to the dilution reservoir 260 to push fluid out of the dilution reservoir 260 for downstream formulation operations. Another fitting 266 connected to a tube which extends to the bottom of the dilution reservoir 260 is used to retrieve diluted product from the dilution reservoir 260 (i.e., product out). The dilution reservoir 260 includes a removable cap 261 that includes corresponding ports or fittings that corresponding to dilution reservoir fittings 262, 264, 266 and tubing which extends into the dilution reservoir 260. Also illustrated in FIG. 4 is a fraction tube holder 268 and can hang from the housing 230 and hold fraction tubes 270. Vials could also be used in place of fraction tubes 270. The fraction tubes 270 are used to collect different fractions obtained after passing through the column 242 and out one of the fraction fittings 272 (four (4) are illustrated but different numbers may be used). In another embodiment, the outputs from the fraction fittings 272 coupled to the fraction collection valve 322 may themselves be used to route products back to the radiosynthesizer 100 for intermediate purification and further reactions. While a compressed, inert gas is used to push fluid out of the dilution reservoir 260 in a preferred embodiment, air could also be used to push fluid out of the dilution reservoir 260. The air could be compressed or it may be contained in a syringe or other pumping device that can be actuated to displace liquid in the dilution reservoir 260. FIGS. 4 and 5 also illustrate liquid sensors 274, 276. These liquid sensors 274, 276 are used to detect the presence (or absence) of liquid in the two outputs from the radiosynthesizer 100. Flexible polymer tubing which connects the respective cassette 102 outputs from the radiosynthesizer 100 is placed in each respective sensor 274, 276. FIG. 6 illustrates a perspective view of the back side of the purification and formulation device 200. Two fans 280, 282 are mounted to the housing 230 and are used to cool the interior electronic and other heat generating components (e.g., UV sensor 300). Electronic input/output ports 284 are located on the back face of the housing 230 and are used to connect to the internal control electronics via the electronics board 310 including, for example, a video I/O port, HPLC pump port, Ethernet port, etc. A switchable power input 286 is provided which connects the purification and formulation device 200 to conventional source of A/C power. Three syringe pump inputs or fittings 288 are also provided which connect to inputs used for the syringe pump (e.g., water, saline solution, eluting solution) as described herein. An inert gas input 290 is located on the back face of the housing 230 and is connected to a source of inert gas such as nitrogen. Three waste ports or fittings 292 are also located on the back face of the housing 230 and connect to flexible tubing (not shown) that leads to waste container(s). FIG. 7 illustrates an interior view of the purification and formulation device 200 with a portion of the housing 230 removed. Located inside the purification and formulation device 200 is a UV detector 300. The UV detector 300 monitors product after traveling through the HPLC column 243. The UV detector 300 can monitor the absorbance of the product(s) passing through polymer tubing at one wavelength between 200 nm and 800 nm. Data from the UV detector 300 is transmitted to the controller 206 and then to the computing device 208 where data can be displayed to the user using the graphical user interface 222 as seen in FIG. 14. A radiation detector 302 is located immediately downstream of the UV detector 300 (with respect to flow path) and measures gamma rays emitted by the decaying radioisotope. A radiation detector 302 amplifier 304 is also located in the purification and formulation device 200 that is used to amplify the signal from the radiation detector 302. Like the data from the UV detector 300, data from the radiation detector 302 is transmitted to the controller 206 and then displayed on the graphical user interface 22 of the computing device 208. Additional detectors could also be incorporated into the flow path. These include, for example, sensors that measure refractive indices, conductivity, and pulsed amperometric detectors (for non-radioactive species). A syringe pump 306 is also located in the purification and formulation device 200 and is used during the formulation operations. The syringe pump 306 is a six (6) port syringe pump that includes an output port, waste port, one input air port (for pushing residual eluting fluid through the lines when the output of the purification and formulation device 200 returns back to the radiosynthesizer 100), and three fluid input ports. One port is coupled via a fluid line to container or reservoir (e.g., Falcon tube) that holds water, another is coupled via a fluid line to a container or reservoir that holds saline, and the final input is coupled via a fluid line to a container or reservoir that holds the eluting fluid. These containers or reservoirs may be located outside of the purification and formulation device 200. An electronics board 310 is also located in the in the purification and formulation device 200 and is used to interface with and control the various sub-systems including the liquid sensors 274, 276, injection valve 240, column selector valve 242, UV detector 300, radiation detector 302, cartridge valve 316, waste valve 318, cleaning valve 320, pressure release valve 326, pressure regulator 400, syringe pump 306, camera 314. Commands and data are communicated with the controller 206 using an Ethernet cable or other communication cable which carries data communications and a video cable that carries the video feed from the camera 314 to the controller 206. The information is read by the software 221 contained on the controller 206 and is then communicated with the computing device 208 for displaying data or assisting the user in making decisions using the GUI 222. A pressure regulator 400 is located in the in the purification and formulation device 200 and is used to regulate the pressure of the inert gas (e.g., nitrogen) that is used to push product from the dilution reservoir 260 during the formulation process as well as being used for cleaning operations. A camera 314 is located in the purification and formulation device 200 and is used to generate live video the dilution reservoir 260 which is then transmitted via the controller 206 to the computing device 208 and user interface 222 for viewing (FIG. 15). The camera 314 is focused on the bottom of the dilution reservoir 260 and indicates to the user how much fluid remains in the reservoir. In some concentration and formulation processes, the user may want to ensure that the dilution reservoir 260 has the appropriate amount of fluid and is not fully exhausted which may damage the trapped probe no the SPE cartridge 252. A series of valves 316, 318, 320, 322, 324 are located in the purification and formulation device 200 and are used to divert various flow paths as described in more detail below. These include a cartridge valve 316 that is used during the formulation process to deliver fluid containing the product from the dilution reservoir 260 to the SPE cartridge 252 or delivery fluid from the syringe pump 306 to the SPE cartridge 252. Waste valve 318 is used to divert fluid from the SPE cartridge 252 either to waste or the final product container 250. Cleaning valve 320 is used to send pressurized inert gas (e.g., nitrogen) through various lines for cleaning the purification and formulation device 200. Fraction collection valve 322 is used to divert fractions to either the fraction tubes 270 to the dilution reservoir 260 for subsequent formulation, or to one of the waste ports 292 on the back of the unit. Pressure release valve 324 is used to permit the passage of air during the filling of the dilution reservoir 260 with liquid. In addition, pressure release valve 324 is also connected to pressurized inert gas which is used to push fluid out of the dilution reservoir 260 and onto the SPE cartridge 252 during the formulation process. FIG. 8 illustrates a schematic layout of the flow paths in the purification and formulation device 200. The formulation subsystem 204 is illustrated by the dashed line in FIG. 8 while the purification subsystem 202 includes the remaining components and processes. As seen in FIG. 8, the injection valve 240 is coupled at one port to the HPLC pump 218 via stainless steel tubing (or other type of tubing) which delivers the carrier/mobile phase used for the HPLC separation process. Multiple different carrier/mobile phases can be run through the HPLC pump 218 as well as mixtures of the same. FIG. 9 illustrates an enlarged view of the injection valve 240. The injection valve 240 is coupled at another port to a first sample input 340 which, in one embodiment, is the output line from the radiosynthesizer 100 from a first cassette 102. The injection valve 240 is coupled at another port to a second sample input 342 which, in one embodiment, is the output line from the radiosynthesizer 100 from a second cassette 102. Liquid sensors 274, 276 are located in respective fluid lines that connect each cassette 102 to the ports on the injection valve 240. A first sample loop 344 is connected to two ports on the injection valve 240 which is used to hold sample from the first sample input 340. A second sample loop 346 is connected to two ports on the injection valve 240 which is used to hold sample from the second sample input 342. The sample loops 344, 346 can contain set volumes of fluid, for example, 5 mL of sample. The two sample loops 344, 346 permit two different purifications runs to be run on the purification and formulation device 200. The injection valve 240 is a two-position valve; when one sample loop (e.g., 344) is in the “load” position, the other loop (e.g., 346) is in the “inject.” Actuation of the injection valve 240 reverses the respective load and injection positions for the sample loops 344, 346. As best seen in FIG. 9, the second loop 346 is in the “load” position while the first loop 344 is in the “inject” position. Note that in an alternative embodiment, sample may be injected into a sample loop 344, 346 manually via input lines 340, 342 using a syringe or the like. The injection valve 240 is also coupled to two waste lines 348, 350 that direct fluid contained therein to waste containers or receptacles 352. The injection valve 240 further includes an output line 354 (e.g., stainless steel) that is connected to the input of the column selector valve 242. The column selector valve 242 is able to connect one of a plurality of columns 243 that may be loaded into the device into the fluid path of the instrument. Still referring to FIG. 8, after fluid passes through the columns, it enters the UV detector 300 followed by the radiation detector 302. The fluid then passes through the cleaning valve 320 that passes the fluid to the fraction collection valve 322 whereby fractions may be collected in fraction tubes 270. Fraction collection valve 322 also includes a waste line 356 that is connected to a waste container or receptacle 357 via waste ports or fittings 292. The fraction collection valve 322 also includes a product line 358 that diverts product to the dilution reservoir 260 which is already pre-loaded or filled with a volume of high-purity water by the user (capacity is 100 ml) into which the product will become diluted. The dilution reservoir 260 also contains an output line 360 that extends into the bottom of the dilution reservoir 260 and is connected at the other end to the cartridge valve 316. The outlet of the cartridge valve 316 is connected to an output line 362 that delivers fluid to the SPE cartridge 252 via port 256. Fluid exits the SPE cartridge 252 and then passes through port 258. The other inlet of the cartridge valve 316 is coupled to the output line 363 of the syringe pump 306. The syringe pump 306 is a six (6) port syringe pump with another port connected to waste line 364 that connects to a waste container or receptacle 358 via waste ports or fittings 292. Another port of the syringe pump 306 connects to an input line 366 that connects to a container or receptacle 368 that contains water. Another port of the syringe pump 306 connects to an input line 370 that connects to a container or receptacle 372 that contains a saline solution. Still another port of the syringe pump 306 connects to an input line 374 that connects to a container or receptacle 376 that eluting fluid (e.g., ethanol or EtOH). Another port of the syringe port 306 is connected to an input line 378 that is open to air 380 that is used to push residual fluid in the fluid carrying lines when product is sent back to the radiosynthesizer 100 for additional unit operations. A sterile air filter 381 may be interposed between the air source 380 and the input to the syringe pump 306. The SPE cartridge 252 is coupled to the waste valve 318 via output line 382. The waste valve 318 includes a waste line 384 that connects to the waste container or receptacle 386 via waste ports or fittings 292. The waste valve 318 is also coupled to an output line 388 that delivers product to the final product container 250. The final product container 250 may include a sterile vial containing a septum which is penetrated by a needle or the like that is secured to a sterile filter 390. The final product container 250 may be contained in a lead pig 251 that limits the emission of radiation. As illustrated in FIG. 8, a vent line 392 is provided along with a sterile filter 394 so that air contained in the final product container 250 can vent out when liquid is delivered to the final product container 250. A pressure regulated inert gas source 400 (e.g., nitrogen) is seen in FIG. 8 connected to the cleaning valve 320 and the pressure release valve 324. As explained herein, the pressurized inert gas source 400 delivers inert gas to the dilution reservoir 260 to push liquid containing the radiopharmaceutical compound into the output line 360 when the pressure release valve 324 is activated. The pressurized inert gas source 400 further is used to clean various fluid lines by passing drying inert gas via cleaning valve 320 during a cleaning procedure. FIGS. 10-13 illustrate the schematic layout of the flow paths used in the four (4) step FORMULATION unit operations. These include FORMULATION: TRAP (FIG. 10); FORMULATION: RINSE (FIG. 11); FORMULATION: ELUTE (FIG. 12); FORMULATION: RECONSTITUTE (FIG. 13). As seen in FIG. 10, in the FORMULATION: TRAP operation, the pressure release valve 324 and the cartridge valve 316 are actuated so that pressurized inert gas from pressure regulator 400 enters the dilution reservoir 260 and pushes liquid containing the radiopharmaceutical compound into the output line 360 where it enters line 362 and into the SPE cartridge 252 whereby the radiopharmaceutical compound becomes “trapped” on the solid phase sorbent material contained therein (e.g., resin). The waste valve 318 is positioned to divert fluid the line 384 and waste container or receptacle 386. Next, with reference to FIG. 11, in the FORMULATION: RINSE operation, the syringe pump 306 is activated to pump water from container or receptacle 368 into output lines 366, 363. The cartridge valve 316 is actuated to pass the water into line 362 where the contents of the SPE cartridge 252 are washed with water. This removes impurities and cleans the fluid lines of organic solvents. The waste valve 318 is positioned to divert fluid the line 384 and waste container or receptacle 386. The volume of rinsing solution may be adjusted using the software 222 on the computing device 208. Next, with reference to FIG. 12, in the FORMULATION: ELUTE operation, the syringe pump 306 is activated to pump eluting fluid (e.g., EtOH) from the container or receptacle 376 into output lines 374, 363. The cartridge valve 316 is actuated to pass the eluting fluid into line 362 where the radiopharmaceutical compound that is trapped in the SPE cartridge 252 elutes off the solid phase sorbent material and into the eluting liquid. The waste valve 318 is actuated to pass this eluting fluid into output line 388 where is passes through the sterile filter 390 and into the final product container 250. Next, with reference to FIG. 13, in the FORMULATION: RECONSTITUTE operation, the syringe pump 306 is activated to pump a saline solution or fluid (e.g., phosphate buffered saline) from the container or receptacle 372 into output lines 370, 363. The cartridge valve 316 is actuated to pass the saline solution or other buffered aqueous solution into line 362 and through the SPE cartridge 252. The waste valve 318 is actuated to pass this saline solution into output line 388 where is passes through the sterile filter 390 and into the final product container 250. This step of the FORMULATION operation allows any residual organic solvents or other unwanted compounds to be diluted to an acceptable level so that the radiopharmaceutical compound is ready for use. Importantly, the syringe pump 306 never encounters any crude or purified product or any of the solvents that may be involved in the TRAP operation. The syringe pump 306 is only programmed to use the fluid path 363 as the output line; it never aspirates fluid through this line. In the embodiment where the output of the product from the output line 388 is returned back to a cartridge 102 of the radiosynthesizer 100, the source of air 380 is pumped by the syringe pump 306 through the sterile filter 381 and into the downstream lines 363, 388 to chase the ethanol or other eluting fluid through the lines. The purification and formulation device 200 may also utilize an automated cleaning operation for both the purification 202 and formulation 204 flow paths. In the purification cleaning operation, a mobile phase is pumped by an HPLC pump 218 through the injector valve 240, column selector valve 242, column(s) 243, UV detector 300, radiation detector 302, cleaning valve 320, fraction collection valve 322, product line 358, waste line 356, and the lines (e.g., four) connected to the fraction containers 270. The mobile phase is collected the dilution reservoir 260 as well as fraction containers 270. After a programmed amount of time, the HPLC pump 218 turns off. The cleaning valve 320 then activates and the pressure regulator 400 outputs compressed inert gas for a programmed time and at a programmed pressure. The fraction collection valve 322 cycles between all outputs (described above) to thoroughly dry the lines. The formulation clean operation cleans the valves and lines used in the FORMULATION operations (FIGS. 10-13). In this process, the dilution reservoir 260 is filled with ethanol (e.g., 100 mL) and the saline line 370 and water line 366 are placed in a waste container. A cleaning solution (e.g., ethanol) in the container or receptacle 376 is also used to complete the cleaning process. The final product line 388 is also placed in a waste container. The SPE cartridge 252 is removed and the input/output fittings 256, 258 are connected together. In the cleaning operation, the syringe pump 306 aspirates the cleaning solution or air to rinse and dry all formulation subsystem 204 fluid paths. Again, both cleaning operations may be performed automatically by the controller 206. FIG. 14 illustrates an example of the GUI 222 that is displayed to the user on the computing device 208. In this example, the user is given current information from the purification subsystem 202 regarding a current purification process being run on the purification and formulation device 200. In this example, the GUI 222 includes a navigation button 500 as well as indicates which subsystem is selected (i.e., purification subsystem 202 or formulation 204 subsystem) by indicator 502 (in FIG. 14 purification is selected but for formulation, indicator 502 will be moved to indicate selection of the formulation operation as seen in FIG. 15). The GUI 222 provides the user with the selected pathway of the fraction collection valve 322 (i.e., product line 358, fraction 270, or waste 356). In this example, the fraction collection valve 322 is diverting product as seen by highlighted selection 504. The GUI 222 further includes a graphing screen 506 that provides live data from the UV detector 300 and/or the radiation detector 302. Various graphing options are provided as seen in option menu 508. These options include AU (for UV graphs), mV (for radiation graphs), output overlay, record, clear, and zero. The GUI 222 also provides a panel of information 510 for control and feedback. This panel 510 displays current operational conditions and configurations of the purification subsystem 202. During operation of the device 200 by the user, various fractions that are separated in the purification subsystem may be selected by the user using a touch button found on the GUI 222 such that the particular fraction is delivered to the fraction collection tubes or vials 270 or delivered as the product to the dilution reservoir 260. This is a process whereby fractions are manually selected by the user. In another alternative embodiment, the software itself can be programmed to automatically select various fractions for shunting to either the collection tubes or vials 270 or the main product line 358. Automated control of the fraction collection valve 322 may look to real time or near real time data generated by the UV detector 300 and radiation detector 302 as well as known elapsed elution times of purified product from the columns 243 to identify suspect or desired fractions of interest. Peak detection algorithms may look at designated time windows and/or set threshold values for the output of UV and/or radiation peaks which can then be used to trigger actuation of the fraction collection valve 322. FIG. 15 illustrates an example of the GUI 222 that is displayed to the user on the computing device 208 when the formulation subsystem 204 is selected by the user as indicated by indicator 502. In FIG. 15, a step selection menu 512 is provided to the user to program and monitor various aspects of the formulation operation. This includes the TRAP operation, the RINSE operation, the ELUTE operation, and the RECONSTITUTE operation. As seen in FIG. 15, the user is provided with a control window 514 that can be used to adjust various operational parameters via edit button 516. GUI 222 also illustrates a live video image 518 of the dilution reservoir 260 taken with camera 314. FIG. 16 illustrates another example of the GUI 222 that is displayed to the user on the computing device 208 and used for data analysis. As seen in FIG. 16 a graph window 520 is provided that is used to display UV data, radiation data, or both. Buttons 522 are selected or deselected to display or hide the desired data. Fractions, if not already named, can be manually named using fraction naming fields 524. The GUI 222 also provides the user the ability to analyze various peaks present in the data fields. For example, the user is able to perform a baseline peak area calculation using baseline button 526. In this example, the radiation detector data is analyzed for peak #1. After identifying the peak and baseline, the software automatically calculates peak area, peak time, and peak height which are displayed in peak output window 528. Peak area may also be expressed as a percentage of the total area, for example, by depressing the area button 529 (which toggles between absolute and percentage values). Point-to-point calculation for peak areas is performed using the point-point button 530 as illustrated in FIG. 16. In peak-to-peak calculation two points are defined that define a line to which the peak is integrated. Data can be exported using export buttons 532 (e.g., batch data or raw .CSV data). Analysis data may be saved or cleared using save button 534 and clear button 536. While embodiments of the present invention have been shown and described, various modifications may be made without departing from the scope of the present invention. For example, while the invention has been described as being usable with the ELIXYS radiosynthesizer available from Sofie Biosciences, the invention is not limited to any particular model or brand. In addition, while the purification and formulation device has been described as including two sample loops, in other embodiments the device may include more than two loops or even a single loop. The invention, therefore, should not be limited except to the following claims and their equivalents.
	description	This application is a U.S. National Stage application under 35 U.S.C. §371, of International Application PCT/JP2012/001953 filed on Mar. 21, 2012, which was published as WO 2013/140445 on Sep. 26, 2013, the disclosure of which is incorporated herein by reference. This invention relates to a radiographic apparatus for obtaining radiological images, and more particularly to a technique for removing scattered radiation using a radiation grid. A conventional radiographic apparatus has a radiation grid for removing scattered radiation in order to prevent scattered radiation from an inspection object from impinging on a flat panel radiation detector (radiation detecting device). The radiation grid is formed of an alternate arrangement of grid foil strips which absorb the scattered radiation and interspacers which transmit the radiation. The grid foil strips are formed of a material such as lead which absorbs radiation, typically X-rays. The interspacers are formed of an intermediate material such as aluminum or an organic material which transmits radiation, typically X-rays. However, when the radiation passes through the interspacers, the radiation (direct radiation) other than the scattered radiation will also be absorbed by the intermediate material. So an air grid, in which the interspacers are made voids for reliably transmitting the radiation (direct radiation) other than the scattered radiation, has been used as radiation grid in recent years. Incidentally, in portions where the direct radiation is blocked by the grid foil strips, foil shadows due to the grid foil strips appear in radiological images. So, Applicant herein has proposed false image removing methods for removing false images resulting from the foil shadows (see Patent Documents 1 and 2, for example). [Patent Document 1] International Publication WO2010-064287 [Patent Document 2] Japanese Unexamined Patent Publication No. 2011-167334 However, the air grid, since the interspacers noted above are voids, easily produces false images due to twisting and bending of the grid foil strips. A false image removing process has been conducted until now with the air grid having grid foil strips arranged synchronously with respective pixels (eg synchronously with every four pixels) forming a radiological image obtained with a flat panel radiation detector (FPD: Flat Panel Detector). It is desired that an appropriate false image removing process is made possible also for the case of grid foil strips in a focusing distance arrangement which is asynchronous. In other words, it is unrealistic to manufacture air grids of sizes compatible and individually synchronized with various FPD pixel sizes. This invention has been made having regard to the state of the art, and its object is to provide a radiographic apparatus which can remove foil shadows in a way to accommodate radiation grids and radiation detecting devices of various pitches or pixel sizes, while taking into consideration twisting and bending of grid foil strips. To fulfill the above object, this invention provides the following construction. A radiographic apparatus according to this invention (radiographic apparatus according to the former invention) is a radiographic apparatus for obtaining a radiological image, comprising a radiation source for emitting radiation; a radiation detecting device for detecting the radiation emitted; and a radiation grid disposed adjacent a detecting plane of the radiation detecting device, and having an arrangement of grid foil strips for absorbing scattered radiation; the radiographic apparatus further comprising an accumulated value calculating device which, in a location where foil shadows by the grid foil strips straddle pixels, identifies the location based on a mutual geometric positional relationship of the radiation source, the radiation detecting device and the radiation grid, and calculates a straddle accumulated value of the foil shadows in the identified location; a radiological image collecting device for collecting an actual radiological image based on radiation detection signals detected in the presence of an inspection object; and a bending constant calculating device for calculating a bending constant which is a constant relating to bending of the grid foil strips in the location where the foil shadows by the grid foil strips straddle the pixels; wherein the radiological image is finally obtained by removing the foil shadows by the grid foil strips based on the accumulated value calculating device, the bending constant calculating device and the radiological image collecting device. A radiographic apparatus according to the latter invention, which is different from the radiographic apparatus according to the former invention, is a radiographic apparatus for obtaining a radiological image, comprising a radiation source for emitting radiation; a radiation detecting device for detecting the radiation emitted; and a radiation grid disposed adjacent a detecting plane of the radiation detecting device, and having an arrangement of grid foil strips for absorbing scattered radiation; the radiographic apparatus further comprising an accumulated value calculating device which, in a location where foil shadows by the grid foil strips straddle pixels, identifies the location based on a mutual geometric positional relationship of the radiation source, the radiation detecting device and the radiation grid, and calculates a straddle accumulated value of the foil shadows in the identified location; a radiological image collecting device for collecting an actual radiological image based on radiation detection signals detected in the presence of an inspection object; and a twisting constant calculating device for calculating a bending constant which is a constant relating to bending of the grid foil strips in the location where the foil shadows by the grid foil strips straddle the pixels; wherein the radiological image is finally obtained by removing the foil shadows by the grid foil strips based on the accumulated value calculating device, the twisting constant calculating device and the radiological image collecting device. The radiographic apparatus according to the former and latter inventions include, besides the radiation source, radiation detecting device and radiation grid, an accumulated value calculating device which, in a location where the foil shadows by the grid foil strips straddle pixels, identifies this location based on a mutual geometric positional relationship of the radiation source, radiation detecting device and radiation grid, and calculates a straddle accumulated value of the foil shadows in the identified location. And the radiological image collecting device is provided for collecting an actual radiological image based on radiation detection signals detected in the presence of an inspection object. A radiological image is finally obtained by removing the foil shadows by the grid foil strips based on the above accumulated value calculating device and the above radiological image collecting device. Even when the foil shadows by the grid foil strips straddle the pixels due to twisting and bending of the grid foil strips, such location is identified based on the mutual geometric positional relationship (that is, geometry) of the radiation source, radiation detecting device and radiation grid, and the straddle accumulated value of the foil shadows in the identified location is calculated. Therefore, even when changes are made in the sizes of the radiation grid and radiation detecting device, the foil shadows will be removed based on the straddle accumulated value. As a result, the foil shadows can be removed taking twisting and bending of the grid foil strips into consideration, and in a way to accommodate radiation grids and radiation detecting devices of various sizes. Twisting or bending of each grid foil strip does not necessarily cause its foil shadow to straddle or cover the pixels. Note that pixels in a location considered likely to be straddled by the foil shadow are recognized from geometry, and a straddle accumulated value in that location is calculated uniformly, regardless of a foil shadow straddle situation. The radiographic apparatus according to the former invention comprises a bending constant calculating device for calculating a bending constant which is a constant relating to bending of the grid foil strips in the location where the foil shadows by the grid foil strips straddle the pixels; wherein the radiological image is finally obtained by removing the foil shadows by the grid foil strips based on the accumulated value calculating device, the bending constant calculating device and the radiological image collecting device. By removing the foil shadows by the grid foil strips using also the bending constant which is a numerical expression of bending, the foil shadows can be removed with increased precision through greater consideration made of the bending of the grid foil strips. The radiographic apparatus according to the latter invention comprises a twisting constant calculating device for calculating a twisting constant which is a constant relating to twisting of the grid foil strips; wherein the radiological image is finally obtained by removing the foil shadows by the grid foil strips based on the accumulated value calculating device, the twisting constant calculating device and the radiological image collecting device. By removing the foil shadows by the grid foil strips using also the twisting constant which is a numerical expression of twisting, the foil shadows can be removed with increased precision through greater consideration made of the twisting of the grid foil strips. The radiographic apparatus according to the former invention (radiographic apparatus with the bending constant calculating device) and the radiographic apparatus according to the latter invention (radiographic apparatus with the twisting constant calculating device) may be combined. That is, the radiological image may be finally obtained by removing the foil shadows by the grid foil strips based on the accumulated value calculating device, the bending constant calculating device, the twisting constant calculating device and the radiological image collecting device. It is preferred that these radiographic apparatus according to this invention comprises an accumulated value multiplying device for multiplying the straddle accumulated value of reference correction data based on X-ray detection signals detected in the absence of the inspection object by a predetermined multiplying factor based on width and pixel size of the foil shadows. By multiplying the straddle accumulated value of the reference correction data by the predetermined multiplying factor, the radiological image without the foil shadows can be obtained in a way to accommodate the radiation grids and radiation detecting devices of various sizes. It is therefore possible to perform an appropriate false image removing process using one radiation grid, without manufacturing a radiation grid according to each radiation detecting device or geometry. The radiographic apparatus according to the former and latter inventions include the accumulated value calculating device which, in a location where the foil shadows by the grid foil strips straddle pixels, identifies this location based on a mutual geometric positional relationship of the radiation source, radiation detecting device and radiation grid, and calculates a straddle accumulated value of the foil shadows in the identified location. Even when the foil shadows by the grid foil strips straddle the pixels due to twisting and bending of the grid foil strips, such location is identified based on geometry and the straddle accumulated value of the foil shadows in the identified location is calculated. Therefore, even when changes are made in the sizes of the radiation grid and radiation detecting device, the foil shadows will be removed based on the straddle accumulated value. As a result, the foil shadows can be removed taking twisting and bending of the grid foil strips into consideration, and in a way to accommodate radiation grids and radiation detecting devices of various sizes. With the radiographic apparatus according to the former invention, by removing the foil shadows by the grid foil strips using also the bending constant which is a numerical expression of bending, the foil shadows can be removed with increased precision through greater consideration made of the bending of the grid foil strips. With the radiographic apparatus according to the latter invention, by removing the foil shadows by the grid foil strips using also the twisting constant which is a numerical expression of twisting, the foil shadows can be removed with increased precision through greater consideration made of the twisting of the grid foil strips. An embodiment of this invention will be described hereinafter with reference to the drawings. FIG. 1 is an outline view and block diagram of an X-ray apparatus according to the embodiment. FIG. 2 is a schematic view of a detecting plane of a flat panel X-ray detector (FPD). FIG. 3 is a schematic view of an X-ray grid. This embodiment will be described taking X-rays as an example of radiation. An X-ray apparatus having a C-arm for application to systems used for cardiovascular diagnosis (CVS: cardiovascular systems) will be described as an example of radiographic apparatus. An air grid with interspacers made voids, which is a focused grid with grid foil strips arranged along rays converging on an X-ray tube, will be described as an example of radiation grid. As shown in FIG. 1, the X-ray apparatus according to this embodiment includes a top board 1 for supporting an inspection object M, an X-ray tube 2 for emitting X-rays, a flat panel X-ray detector (hereinafter abbreviated as “FPD”) 3 for detecting the X-rays emitted, and an X-ray grid 4 disposed adjacent a detecting plane of the FPD 3 and having an arrangement of grid foil strips 4a (see FIG. 3, for example) for absorbing scattered X-rays. The X-ray tube 2 corresponds to the radiation source in this invention. The flat panel X-ray detector (FPD) 3 corresponds to the radiation detecting device in this invention. The X-ray grid 4 corresponds to the radiation grid in this invention. In addition, the X-ray apparatus includes a C-arm 5 which holds the X-ray tube 2 at one end thereof, and holds the FPD 3 along with the X-ray grid 4 at the other end. In FIG. 1, the C-arm 5 is formed in the shape of a curve in the body axis direction of the inspection object M. The C-arm 5 is rotatable along the C-arm 5 itself and about a rotation center axis perpendicular to the body axis of the inspection object M, whereby the X-ray tube 2, FPD 3 and X-ray grid 4 held by the C-arm 5 can also be rotated in the same direction. Further, the C-arm 5 is rotatable about a rotation center axis perpendicular to the body axis, whereby the X-ray tube 2, FPD 3 and X-ray grid 4 can also be rotated in the same direction. Specifically, the C-arm 5 is held by a base block 6 fixed to a floor, through a strut 7 and an arm holder 8. The strut 7 is rotatable about a vertical axis relative to the base block 6, and with this rotation the X-ray tube 2, FPD 3 and X-ray grid 4 can also be rotated in the same direction together with the C-arm 5 held by the strut 7. With the arm holder 8 held by the strut 7 to be rotatable about the body axis of the inspection object M, the X-ray tube 2, FPD 3 and X-ray grid 4 can also be rotated in the same direction together with the C-arm 5 held by the arm holder 8. With the C-arm 5 held by the arm holder 8 to be rotatable about the rotation center axis, the X-ray tube 2, FPD 3 and X-ray grid 4 can also be rotated in the same direction together with the C-arm 5. Further, a construction may be provided for moving the FPD 3 to and fro along an emission axis of X-rays linking the X-ray tube 2 and FPD 3, or to and fro in a focus line direction perpendicular to the emission axis. Even under a condition that the positional relationship between the X-ray tube 2, FPD 3 and X-ray grid 4 should be constant, shifting may be caused, for example by rotation of the C-arm 5, in the positional relationship between the X-ray tube 2, FPD 3 and X-ray grid 4 (transverse focal shift amount=Xf to be described hereinafter). Further, the X-ray apparatus includes an image processor 11 for carrying out various image processes based on X-ray detection signals detected by the FPD 3, a memory unit 12 for writing and storing reference correction data obtained in advance of X-raying and data of each image obtained by the image processor 11, for example, an input unit 13 for inputting data and commands, a display unit 14 for displaying images obtained by the image processor 11, and a controller 15 for performing overall control of these components. In addition, a high voltage generator, for example, is provided for generating high voltage and applying tube current and tube voltage to the X-ray tube 2. However, this does not constitute the characterizing part of this invention or is not a component relating to the characterizing part, and is therefore omitted from the drawings. The memory unit 12, through the controller 15, writes and stores the reference correction data and data of each image obtained by the image processor 11, which are read as appropriate if necessary, and through the controller 15, these data are fed to and displayed on the display unit 14. The memory unit 12 is formed of a storage medium represented by a ROM (Read-only Memory), a RAM (Random-Access Memory), a hard disk and so on. The input unit 13 feeds into the controller 15 the data and commands inputted by the operator. The input unit 13 is formed of a pointing device represented by a mouse, a keyboard, a joystick, a trackball, a touch panel and so on. The display unit 14 is formed of a monitor. The image processor 11 and controller 15 are formed of a central processing unit (CPU) and others. The data of each image obtained by the image processor 11 is written and stored through the controller 15 in the memory unit 12, or is fed into and displayed on the display unit 14. A specific construction of the image processor 11 will be described in detail hereinafter. The FPD 3, as shown in FIG. 2, has a plurality of detecting elements d sensitive to X-rays and arranged in a two-dimensional matrix form on the detecting plane thereof. The detecting elements d detect X-rays by converting the X-rays transmitted through the inspection object M into X-ray detection signals (electric signals) to be stored once, and reading the X-ray detection signals stored. The X-ray detection signal detected by each detecting element d is converted into a pixel value corresponding to the X-ray detection signal. An X-ray image is outputted by allotting the pixel values to pixels corresponding to positions of the detecting elements d. The X-ray image is fed to the image processor 11. The X-ray grid 4, as shown in FIG. 3, has an alternate arrangement of grid foil strips 4a for absorbing scattered X-rays and interspacers 4b for transmitting X-rays. The grid foil strips 4a and interspacers 4b are covered by grid covers 4c located on an X-ray incidence plane and on an opposite plane with the grid foil strips 4a and interspacers 4b in between. In order to clarify illustration of the grid foil strips 4a, the grid covers 4c are shown in two-dot chain lines, and other details of the X-ray grid 4 (eg a structure for supporting the grid foil strips 4a) are not shown. The grid foil strips 4a correspond to the grid foil strips in this invention. As shown in FIG. 3, the X-ray grid 4 is placed to have the respective grid foil strips 4a arranged parallel to the detecting plane of the FPD 3. In this embodiment, the interspacers 4b are voids, and the X-ray grid 4 is also an air grid. The grid foil strips 4a are not limited to a particular material as long as it absorbs radiation represented by X-rays, such as lead. In this embodiment, it is a focused grid having the grid foil strips 4a arranged along rays converging on the X-ray tube 2 (see FIG. 1), but for expediency of illustration, the respective grid foil strips 4a are arranged parallel in FIG. 3. Assuming that each pixel size is ΔX as shown in FIG. 3, the grid foil strips 4a are arranged synchronously with the respective pixels in this embodiment for facility of understanding. That is, in FIGS. 5 and 7 where distortion (bending and twisting) information on the grid foil strips 4a is collected, the grid foil strips 4a are arranged synchronously with every four pixels. Therefore, with the grid foil strips 4a absorbing X-rays, foil shadows are produced on the FPD 3 and the foil shadows appear on X-ray images, but the grid foil strips 4a are arranged so that the foil shadows may appear synchronously with the respective pixels. Next, the image processor and a flow of a series of image processes will be described with reference to FIGS. 4-11. FIG. 4 is a block diagram of a specific image processor according to the embodiment. FIG. 5 is a schematic view showing a positional relationship when obtaining a bending constant. FIG. 6 is a schematic view when the bending constant is applied to actual radiography. FIG. 7 is a schematic view showing a positional relationship when obtaining a twisting constant. FIG. 8 is a schematic view when the twisting constant is applied to actual radiography. FIG. 9 is a schematic view schematically showing calculation of the bending constant. FIG. 10 is a schematic view of a profile of a straddle accumulated value of reference correction data. FIG. 11 is a schematic view illustrating multiplication of the straddle accumulated value of reference correction data by a predetermined multiplying factor. The image processor 11, as shown in FIG. 4, includes a correction data collecting unit 31, a radiological image collecting unit 32, a first accumulated value calculating unit 33, an accumulated value multiplying unit 34, a corresponding corrected image calculating unit 35, a foil shadow aligned image generating unit 36, a lowpass filter (hereinafter abbreviated as “LPF”) 37, a second accumulated value calculating unit 38, a false image removing foil shadow image generating unit 39 and a false image removed image generating unit 40. The correction data collecting unit 31 has a bending constant calculating unit 31a and a twisting constant calculating unit 31b. The bending constant calculating unit 31a corresponds to the bending constant calculating device in this invention. The twisting constant calculating unit 31b corresponds to the twisting constant calculating device in this invention. The radiological image collecting unit 32 corresponds to the radiological image collecting device in this invention. The first accumulated value calculating unit 33 and second accumulated value calculating unit 38 correspond to the accumulated value calculating device in this invention. The accumulated value multiplying unit 34 corresponds to the accumulated value multiplying device in this invention. Here, a home position (HP: Home Position) is a converging position of foil inclinations in an ideal case where the grid foil strips 4a are free from bending and twisting, which refers to a focal position HP located on a center line of the FPD 3 and X-ray grid 4 as shown in FIGS. 5-8 and located at a distance of f0 from the X-ray grid 4. SID is, when a perpendicular is drawn from the focal position on the X-ray tube 2 to the FPD 3, a distance from the focal position in the perpendicular direction to the FPD 3 (SID: Source Image Distance), and f0 is a distance (focusing distance) between the home position HP and a center plane of the X-ray grid 4. As shown in FIGS. 5-8, the coordinates of the home position HP are set to (0, 0). When, as shown in FIGS. 6 and 8, a transverse focal shift amount of the X-ray tube 2 (see FIG. 1) (focal shift amount along the installation plane direction of the FPD 3 and X-ray grid 4 from the home position HP) is Xf and a longitudinal focal shift amount (focal shift amount along the perpendicular direction from the home position HP) is dr, the coordinates of an actual radiographic focus will become (Xf, dr). When it is assumed that a grid foil strip 4a serving as reference is located on the center line of the FPD 3 and X-ray grid 4, and that this grid foil strip 4a has shifted as a result of grid attachment and detachment, for example, the shift amount is set to Xg. Since the longitudinal shift amount dr is set as a use situation of the X-ray apparatus, it is an amount readable from the apparatus and is known. The following description will be made on the assumption that the transverse focal shift amount Xf and grid shift amount Xg at the time of actual radiography are also known from a marker process, a correlation process by foil shadows, or other processes. The correction data collecting unit 31 collects reference correction data before shipment. In particular, the bending constant calculating unit 31a calculates a bending constant, and a twisting constant calculating unit 31b calculates reference correction data having twisting constant information. The radiological image collecting unit 32, when an actual radiological image is labeled I as shown in FIG. 4, collects the actual radiological image based on X-ray detection signals detected in the presence of the inspection object M (see FIG. 1). At the time of actual radiography, as shown in FIGS. 6 and 8, each pixel size is ΔX′ and the distance between the detecting plane of the FPD 3 and the center plane of the X-ray grid 4 is G′. When collecting the reference correction data in advance of X-raying (eg before shipment of the X-ray grid 4), as shown in FIGS. 5 and 7, with each pixel size being ΔX0 and the distance between the detecting plane of the FPD 3 and the center plane of the X-ray grid 4 being G0, it is not absolutely necessary that ΔX0=ΔX′ and G0=G′. Even ΔX0≠ΔX′ and G0≠G′ will do. In other words, as noted in the Technical Problem section hereof, the following image processing is applicable in a way to accommodate radiation grids and radiation detecting devices of various sizes. The actual radiological image I collected by the radiological image collecting unit 32 is fed to the foil shadow aligned image generating unit 36. The bending constant calculating unit 31a, when the bending constant is labeled δtn as shown in FIG. 4, calculates bending constant δtn which is a constant relating to bending of the grid foil strips 4a in locations where the foil shadows cast from the grid foil strips 4a straddle pixels. Specifically, a correction data collecting device having a layout as shown in FIG. 5 is used to move the X-ray grid 4 at predetermined intervals (eg about 20 μm) along the installation plane of the X-ray grid 4 in the absence of an inspection object, and bending constant δtn is obtained by collecting signal strengths of the straddle position pixels. When obtaining reference correction data having twisting constant δθn information, a correction data collecting device having a layout as shown in FIG. 7 is used. When collecting the reference correction data, in the absence of an inspection object, the focus is moved from the home position HP at predetermined intervals (eg about 1 mm) along a focus line Lc as shown in FIG. 7, and an X-ray image is collected for every focal position as reference correction data through the FPD 3. At this time, the reference correction data is collected using a device other than the X-ray apparatus used for actual radiography, which is a correction data collecting device which does not easily produce shifting of the positional relationship. Of course, when an X-ray apparatus of the type which does not easily produce focal shifting from one radiography to another, the same X-ray apparatus may be used to collect the reference correction data. Returning to the description of FIG. 5, in the state of the focus of the X-ray tube 2 (see FIG. 1) set to the home position HP, X-ray detection signals are collected, respectively, by moving the X-ray grid 4 at predetermined intervals along the installation plane of the X-ray grid 4 as described above. The grid foil strip 4a used as reference is an n0th foil strip which is the n0th in order, and a target grid foil strip 4a is an nth foil strip which is the nth in order. At this time, each signal strength of the X-ray detection signals in a plurality of pixels straddling the foil shadow by the nth target foil strip is collected as signal strength of a straddle position pixel. Although the n0th foil strip serving as the reference is the grid foil strip 4a located on the center line in FIGS. 5-8, this is not limitative and, for example, a grid foil strip 4a located at an end may be used as the foil strip serving as the reference. To describe that a foil shadow straddles two pixels in the case of FIG. 5, signal strengths of the two pixels straddling the foil shadow by the nth foil strip serving as the target are collected, respectively, as signal strengths of the straddle position pixels as shown in FIG. 9(b). In FIG. 9(b), the horizontal axis represents amount of movement of the X-ray grid 4, and the vertical axis represents signal strength. When there is no bending of the grid foil strips 4a, and assuming that the boundary between the two pixels straddles the center of the foil shadow, the position where the signal strengths of the straddle position pixels intersect each other should be located at the boundary between the two pixels. In practice, however, due to bending of the grid foil strips 4a, as shown in FIG. 9(b), the signal strengths of the straddle position pixels intersect each other in a position deviating from the boundary (shown in a dotted line) between the two pixels. This deviation of the intersecting position from the boundary between the two pixels is defined as bending constant (see also FIG. 9(a)) δtn. FIG. 9(a) is a schematic view of the foil shadow by the nth foil strip projected in a bent state on the pixels. This bending constant δtn is obtained for each pixel row and each grid foil strip 4a. The bending constant described here and also the twisting constant described hereinafter, of course, differ from foil strip to foil strip, and vary also for each position in the foil running direction (row), but this embodiment is described as treatment in a fixed row position. In this way, the bending constant calculating unit 31a calculates bending constant δtn as shown in FIG. 4. The bending constant δtn calculated by the bending constant calculating unit 31a is fed to the first accumulated value calculating unit 33 and second accumulated value calculating unit 38. When applying bending constant δtn to actual radiography, foil shadow pixels at the time of actual radiography are identified by calculating Kn and also calculating Dn as shown in FIG. 6. In FIGS. 6 and 8, the X-ray grid 4 shifted at the time of actual radiography (grid arrangement at the time of radiography) is indicated in two-dot chain lines, to show it in distinction from the X-ray grid 4 at the time of calculation (grid arrangement at the time of calculation) which is indicated in solid lines. When the focus shifted by grid shift amount Xg from the actual radiographic focus (Xf, dr) is a calculated focus, the coordinates of the calculated focus are (Xf−Xg, dr) as shown in FIGS. 6 and 8. When the distance between the nth target foil strip in the grid arrangement at the time of calculation and the center line (of the FPD 3 and X-ray grid 4) is Kn, distance Ku (that is, position from the center line of the nth foil strip) can be derived from the following equation (1):Kn=(n−n0)·P+δtn (1) where P is an ideal foil strip pitch, and P is known. Bending constant δtn also has already been obtained. Therefore, distance Kn can be obtained since the right-hand side of equation (1) above is known. When, among the rays from the calculated focus (Xf−Xg, dr), the distance of a position projected on the FPD 3 through the nth foil strip in the grid arrangement at the time of calculation and the center line is Dn, distance Dn can be obtained based on the geometric positional relationship of the following equation (2):Dn=Kn−{(Xf−Xg−Kn)·G′/(f0+dr)} (2) As described hereinbefore, f0 is the distance (focusing distance) between the home position HP and the center plane of the X-ray grid 4, and f0 is known. It is premised that the longitudinal focal shift amount dr, and the distance G′ between the detecting plane of the FPD 3 and the center plane of the X-ray grid 4, are known, and that each of the shift amounts Xf and Xg is known as noted above. The distance Kn which is the position from the center line of the nth foil strip has already been derived from equation (1) above. Therefore, distance Dn can be obtained since the right-hand side of equation (2) above is known. Next, foil shadow pixels are identified as locations where the foil shadows of the grid foil strips 4a straddle pixels, using the distance Dn derived from equation (2) above and pixel size ΔX′ at the time of actual radiography. Specifically, the distance Dn is divided by pixel size ΔX′, and foil shadow pixels are identified from the integer portion of the division result Dn/ΔX′, and detailed straddle positions from the fractional portion thereof. A straddle accumulated value of the nth foil strip at the time of radiography is obtained by subtracting the signal strengths of the identified foil shadow pixels from the signal strengths at the time when it is assumed that there is no foil shadow. By obtaining this for each pixel row and each grid foil strip 4a, the first accumulated value calculating unit 33 can calculate a straddle accumulated value of reference correction data described hereinafter, and the second accumulated value calculating unit 38 can calculate a straddle accumulated value of a foil shadow enhanced image based on an actual radiological image described hereinafter. The first accumulated value calculating unit 33, as shown in FIG. 4, in a location where foil shadows of the grid foil strips 4a straddle pixels, identifies this location based on a mutual geometric positional relationship (that is, geometry) of the X-ray tube 2, FPD 3 and X-ray grid 4 (see FIG. 1 for all), and calculates a straddle accumulated value of the foil shadows in the identified location. Similarly, the second accumulated value calculating unit 38 identifies the location based on geometry, and calculates a straddle accumulated value of the foil shadows in the identified location. There are two straddle accumulated values, which are a straddle accumulated value of the reference correction data described hereinafter and a straddle accumulated value of a foil shadow enhanced image based on an actual radiological image described hereinafter. As noted above, the first accumulated value calculating unit 33 calculates the straddle accumulated value of the reference correction data, and the second accumulated value calculating unit 38 calculates the straddle accumulated value of the foil shadow enhanced image. As noted hereinbefore, twisting or bending of each grid foil strip 4a does not necessarily cause its foil shadow to straddle or cover the pixels. Depending on a twist or bend situation, the foil shadow may not cover even one pixel but may cover other pixels (eg adjacent pixels). In that case, pixels in a location considered likely to be straddled by the foil shadow are recognized from the mutual geometric positional relationship (that is, geometry) of the X-ray tube 2, FPD 3 and X-ray grid 4, and straddle accumulated values in that location are calculated uniformly, regardless of a foil shadow straddle situation. The straddle accumulated value of the reference correction data based on calculation by the first accumulated value calculating unit 33 will be described in detail. A profile of the straddle accumulated value of the reference correction data is obtained as shown in FIG. 10. In FIG. 10, the horizontal axis represents amounts of focal movement, and the vertical axis represents ratios between the denominator which is a signal strength when assuming that there is no foil shadow, and the numerator which is a straddle accumulated value of the nth foil strip, thereby creating a profile of the straddle accumulated value of the nth foil strip. When there is no twist in the grid foil strips 4a, the profile is as shown in the dotted lines in FIG. 10. In practice, however, due to twisting occurring to the grid foil strips 4a, the profile becomes what is shown in the solid lines in FIG. 10. Then, the twisting constant calculating unit 31b calculates a twisting constant. The twisting constant calculating unit 31b, when the twisting constant is labeled δθn as shown in FIG. 4, calculates twisting constant δθn which is a constant relating to twisting of the grid foil strips 4a. When calculating twisting constant δθn as above, the correction data collecting device shown in FIG. 7 is used, which is the same as what is shown in FIG. 5. When, as shown in FIG. 7, there is a twist at twist angle δθn from an ideal angle of the nth foil strip, a ray forming twist angle δθn with a ray passing through the nth foil strip serving as the target is defined from among the rays from the home position HP. A profile of the straddle accumulated value based on the data at the focus (indicated by a void square: “□” in FIG. 7) at which this defined ray and the focus line Lc meet becomes a profile as shown in the solid lines in FIG. 10. Therefore, a shifting in an amount of focal movement from the profile of the nth ideal foil strip as shown in the dotted lines in FIG. 10 can be detected by collecting data in the state of the grid foil strip 4a being twisted, and creating the profile (solid lines in FIG. 10) based thereon. Since the shifting at this time is an amount (indicated “∝ sin (δθn)” in FIG. 10) proportional to sin (δθn), it is possible to obtain twisting constant δθn directly, but a straddle accumulated value for each sin (δθn) is stored in order to reduce the amount of calculation. This is called reference correction data having twisting constant δθn information. Each mark in FIG. 7, each mark in FIG. 8 and each mark in FIG. 10 are unified (for example, focus □ where the ray and the focus line Lc meet is common to FIGS. 7, 8 and 10). In this way, the twisting constant calculating unit 31b calculates twisting constant δθn (reference correction data) as shown in FIG. 4. The twisting constant δθn (reference correction data) calculated by the twisting constant calculating unit 31b is fed to the first accumulated value calculating unit 33. When applying twisting constant δθn (reference correction data) to actual radiography, as shown in FIG. 8, Kn is calculated, and further Dn is calculated, thereby to determine a point of intersection with the focus line Lc (indicated by a black square: “▪” in FIG. 8). Since this intersection is not necessarily in agreement with the focal position of the time when the reference correction data is collected, an nth straddle accumulated value of a corresponding corrected image is obtained, for example, by making linear interpolation through weighting correction from straddle accumulated values in two adjacent focal positions. Therefore, when located at midpoint between focal positions, linear interpolation may be carried out with the same weighting of straddle accumulated values in two adjacent focal positions. The method of calculating Kn and Dn has been described in FIG. 6, and its description is omitted here. When the intersection (▪ in FIG. 8) with the focus line Lc is located in the peak position of the profile, extrapolation may be carried out from opposite sides. Thus, the first accumulated value calculating unit 33 calculates a straddle accumulated value of the reference correction data, using the bending constant δtn calculated by the bending constant calculating unit 31a and the twisting constant δθn calculated by the twisting constant calculating unit 31b (reference correction data). As described above, the straddle accumulated value of the nth foil strip of the reference correction data is obtained by subtracting the signal strengths in the identified foil shadow pixels in the reference correction data from the signal strengths in the reference correction data of the time when assuming that there is no foil shadow. Since calculation is carried out for the foil position reflecting bending constant δtn at this time, the straddle accumulated value of the reference correction data can be calculated accurately. In this way, the first accumulated value calculating unit 33 calculates the straddle accumulated value of the reference correction data as shown in FIG. 4. The straddle accumulated value of the reference correction data calculated by the first accumulated value calculating unit 33 is fed to the accumulated value multiplying unit 34 and corresponding corrected image calculating unit 35. The accumulated value multiplying unit 34, as shown in FIG. 4, multiplies the straddle accumulated value of the reference correction data by a predetermined multiplying factor based on the width and pixel size of the foil shadow. As described hereinbefore, pixel size ΔX0 at the time of collecting the reference correction data and pixel size ΔX′ at the time of actual radiography are not necessarily equal. In FIG. 11, the width of the foil shadow is fixed to W (see FIG. 11(b)) for simplification. Assuming that ΔX0=150 μm (=0.15 mm), ΔX′=200 μm (=0.2 mm) and W=0.034 mm, the peak value of the profile of the straddle accumulated values, as shown in FIG. 11(a), is 0.77 (ΔX0−W)/ΔX0=(0.2−0.034)/0.2) at the time of collecting the reference correction data, and is 0.83 (=(ΔX′−W)/ΔX′=(0.15−0.034)/0.15) at the time of actual radiography. Each mark in FIG. 7, each mark in FIG. 8, each mark in FIG. 10 and each mark in FIG. 11(a) are unified (for example, focus □ where the ray and the focus line Lc meet is common to FIGS. 7, 8, 10 and 11(a), and intersection ▪ with the focus line Lc is common to FIGS. 8 and 11(a)). In this way, the accumulated value multiplying unit 34, as shown in FIG. 4, multiplies the straddle accumulated value of the reference correction data by the multiplying factor (0.83/0.77 here). The straddle accumulated value of the reference correction data multiplied by the accumulated value multiplying unit 34 is fed to the corresponding corrected image calculating unit 35. Based on the straddle accumulated value of the reference correction data multiplied by the accumulated value multiplying unit 34, the corresponding corrected image calculating unit 35 determines a ray from the calculated focus shifted from an actual radiographic locus by the grid shift amount Xg (see FIGS. 6 and 8), and obtains a corresponding corrected image corresponding to that ray. As described also in FIGS. 6 and 8, the coordinates of the calculated focus shifted from the actual radiographic focus (Xf, dr) by the grid shift amount Xg are (Xf−Xg, dr). At this time, if reference correction data at the intersection (indicated by black square: “▪” in FIG. 8) of the ray passing through the nth foil strip, among the rays from the calculated focus (Xf−Xg, dr), and the focus line Lc (see FIGS. 6 and 8), a corresponding corrected image of that time can be obtained. When the intersection at which the ray concerned and the focus line Lc meet is in agreement with the focal position of the time when the reference correction data is collected, the reference correction data (X-ray image) in that focal position may serve as it is as the corresponding corrected image. However, the intersection at which the ray concerned and the focus line Lc meet is not necessarily in agreement with the focal position of the time when the reference correction data is collected. In that case, a corresponding corrected image can be obtained by making weighting correction using reference correction data (X-ray images) in two focal positions closest to the intersection (that is, adjoining each other) at which the ray concerned and the focus line Lc meet, respectively. A corresponding corrected image can be obtained, for example, by allotting to each pixel a pixel value (value of X-ray detection signal) which is a sum of a product of reference correction data in one of the adjacent focal positions and a weighting function of that time, and a product of reference correction data in the other focal position and the weighting function of that time. In this way, the corresponding corrected image calculating unit 35 calculates a corresponding corrected image as shown in FIG. 4. Since the straddle accumulated value of the reference correction data is already obtained by the first accumulated value calculating unit 33, assuming that the accumulated value in the corresponding corrected image (the nth straddle accumulated value of the above corresponding corrected image) is Csum, the accumulated value Csum in the corresponding corrected image is fed to the false image removing foil shadow image generating unit 39. When a foil shadow aligned image is labeled G as shown in FIG. 4, the foil shadow aligned image generating unit 36 generates the foil shadow aligned image G, which is a radiological image showing aligned foil shadows, by sliding the radiological image I in the direction of arrangement of the grip foil strips 4a (lateral direction in FIG. 3). The foil shadow aligned image G can be generated by sliding the radiography image I with information on the inspection object, using the grid shift amount Xg described above. When generating a foil shadow aligned image G with higher precision, a precise foil shadow aligned image G can be generated by obtaining a shift amount (grid shift amount Xg) for each pixel row and sliding the radiological image I for each pixel row. There remains a shift amount in the extending direction of the grid foil strips 4a (vertical direction in FIG. 3), but such shift amount is slight and thus negligible. In this way, the foil shadow aligned image generating unit 36 generates the foil shadow aligned image G as shown in FIG. 4. The foil shadow aligned image G generated by the foil shadow aligned image generating unit 36 is fed to the LPF 37 and false image removed image generating unit 40. The LPF 37, when the foil shadow enhanced image is labeled E as shown in FIG. 4, passes low-pass areas in the longitudinal direction of the grid foil strips 4a (vertical direction in FIG. 3) in order to generate the foil shadow enhanced image E enhancing the foil shadows in the foil shadow aligned image G and with the information on the inspection object M (see FIG. 1) removed therefrom. The foil shadow enhanced image F generated by the LPF 37 is fed to the second accumulated value calculating unit 38. The second accumulated value calculating unit 38, when a straddle accumulated value of the foil shadow enhanced image E is labeled Esum as shown in FIG. 4, calculates the accumulated value Esum. The accumulated value Esum is fed to the false image removing foil shadow image generating unit 39. The false image removing foil shadow image generating unit 39, when a false image removing foil shadow image is labeled Cor as shown in FIG. 4, can generate the false image removing foil shadow image Cor for removing false images resulting from the foil shadows based on the accumulated values Esum and sum (Cor=E·Csum/Esum). In this way, the false image removing foil shadow image generating unit 39 generates the false image removing foil shadow image Cor as shown in FIG. 4. The false image removing foil shadow image Cor generated by the false image removing foil shadow image generating unit 39 is fed to the false image removed image generating unit 40. The false image removed image generating unit 40, when an X-ray image finally obtained by having the foil shadows removed is labeled Iafter as shown in FIG. 4, generates the false image removed image without the foil shadows by the grid foil strips 4a, based on the false image removing foil shadow image Cor. And the false image removed image generated by the false image removed image generating unit 40 is finally obtained as X-ray image Iafter. The X-ray image Iafter can be obtained by dividing the foil shadow aligned image G by the false image removing foil shadow image Cor for each pixel (Iafter=G/Cor). Since the interspacers are voids in the case of an air grid, the contrast between pixels straddled by the foil shadows and pixels not straddled is strong, and the false images are conspicuous. The problem addressed by the invention is solvable by applying the image processor and the flow of a series of image processes described above to the air grid. The X-ray apparatus according to this embodiment includes, besides the X-ray tube 2, FPD 3 and X-ray grid 4, the first accumulated value calculating unit 33 and second accumulated value calculating unit 38 which, in a location where the foil shadows by the grid foil strips 4a straddle pixels, identify this location based on a mutual geometric positional relationship of the X-ray tube 2, FPD 3 and X-ray grid 4, and calculate straddle accumulated values of the foil shadows in the identified location. And the radiological image collecting unit 32 is provided for collecting an actual radiological image based on X-ray detection signals detected in the presence of an inspection object M. An X-ray image is finally obtained by removing the foil shadows by the grid foil strips 4a based on the above first and second accumulated value calculating units 33 and 38 and the above radiological image collecting unit 32. Even when the foil shadows by the grid foil strips 4a straddle the pixels due to twisting and bending of the grid foil strips 4a, such location is identified based on the mutual geometric positional relationship (that is, geometry) of the X-ray tube 2, FPD 3 and X-ray grid 4, and the straddle accumulated values of the foil shadows in the identified location are calculated. Therefore, even when changes are made in the sizes of the X-ray grid 4 and FPD 3, the foil shadows will be removed based on the straddle accumulated values. As a result, the foil shadows can be removed taking twisting and bending of the grid foil strips 4a into consideration, and in a way to accommodate X-ray grids 4 and FPDs 3 of various sizes. In this embodiment, it is preferred to provide the bending constant calculating unit 31a for calculating a bending constant which is a constant relating to bending of the grid foil strips 4a in the location where the foil shadows by the grid foil strips 4a straddle the pixels, wherein the X-ray image is finally obtained by removing the foil shadows by the grid foil strips 4a based on the first and second accumulated value calculating units 33 and 38, the bending constant calculating unit 31a and the radiological image collecting unit 32. By removing the foil shadows by the grid foil strips 4a using also the bending constant which is a numerical expression of bending, the foil shadows can be removed with increased precision through greater consideration made of the bending of the grid foil strips 4a. In this embodiment, it is preferred to provide the twisting constant calculating unit 31b for calculating a twisting constant which is a constant relating to twisting of the grid foil strips 4a, wherein the X-ray image is finally obtained by removing the foil shadows by the grid foil strips 4a based on the first and second accumulated value calculating units 33 and 38, the twisting constant calculating unit 31b and the radiological image collecting unit 32. By removing the foil shadows by the grid foil strips 4a using also the twisting constant which is a numerical expression of twisting, the foil shadows can be removed with increased precision through greater consideration made of the twisting of the grid foil strips 4a. In this embodiment, it is preferred to provide the accumulated value multiplying unit 34 for multiplying a straddle accumulated value of reference correction data based on X-ray detection signals detected in the absence of the inspection object by a predetermined multiplying factor based on the width and pixel size of the foil shadows. By multiplying the straddle accumulated value of the reference correction data by the predetermined multiplying factor, the X-ray image without the foil shadows can be obtained in a way to accommodate the X-ray grids 4 and FPD 3 of various sizes. It is therefore possible to perform an appropriate false image removing process using one X-ray grid, without manufacturing an X-ray grid according to each FPD or geometry. This invention is not limited to the foregoing embodiment, but may be modified as follows: (1) The foregoing embodiment has been described taking X-rays as an example of radiation. However, the invention is applicable to radiation other than X-rays (such as gamma rays). (2) In the foregoing embodiment, the X-ray apparatus is an apparatus having a C-arm for application to CVS systems, but this is not limitative. For example, the apparatus may be constructed like a nondestructive testing apparatus for industrial use which conducts radiography of an object (in this case, a subject tested) conveyed on a belt, or may be constructed like an X-ray CT apparatus for medical use. (3) In the foregoing embodiment, an air grid is employed as radiation grid, but this is not limitative. The grid may have, in place of the voids, an intermediate material such as aluminum or organic substance which transmits radiation represented by X-rays. Further, the grid may be a cross grid. In the case of a cross grid, grid shifting is less likely to occur than the air grid with grid foil strips extending only in one direction, but of course, it is applicable. In this case, the direction of shifting may be determined with an extension from one direction to two directions. (4) The foregoing embodiment provides a focused grid, but the invention is applicable also to a grid of parallel arrangement. (5) The foregoing embodiment has been described in relation to a grid synchronous (synchronous grid) with the pixels (see ΔX0 in FIGS. 5 and 7) of the FPD 3 at the time of correction data collection, but the invention may be applied to an asynchronous grid. In the case of a grid other than the air grid, it may be applied to a grid having a construction in which a plurality of grid foil strips are juxtaposed for one pixel. (6) The foregoing embodiment provides the bending constant calculating device (bending constant calculating unit 31a in the embodiment) and the twisting constant calculating device (twisting constant calculating unit 31b in the embodiment) in order to calculate a bending constant and a twisting constant which are numerical expressions of bending and twisting, respectively. If there is little influence of bending and twisting, only the accumulated value calculating devices (first and second accumulated value calculating units 33 and 38) may be provided for calculating straddle accumulated values based on geometry. (7) In the foregoing embodiment, the straddle accumulated value of reference correction data is multiplied by the predetermined multiplying factor based on the width and pixel size of the foil shadows. However, where the X-ray grid 4 and FPD 3 are the same size, or geometry is the same, it is not absolutely necessary to multiply the straddle accumulated value of reference correction data by the predetermined multiplying factor. (8) In the foregoing embodiment, the straddle accumulated values are data relating to the reference correction data and the foil shadow enhanced image, but this is not (imitative. For example, straddle accumulated values relating to the radiological image may be obtained. 2 . . . X-ray tube 3 . . . flat panel X-ray detector (FPD) 4 . . . X-ray grid 4a . . . grid foil strips 31a . . . bending constant calculating unit 31b . . . twisting constant calculating unit 32 . . . radiological image collecting unit 33 . . . first accumulated value calculating unit 34 . . . accumulated value multiplying unit 38 . . . second accumulated value calculating unit M . . . inspection object
	claims	1. An electron-beam calibration method for an electron-beam system including at least an electron source for emitting an electron beam, a stage for mounting a sample thereon, deflection means, an objective lens, and a calibrating mark, said electron-beam calibration method comprising the steps of:scanning said electron beam on said calibrating mark in a first direction parallel to a one-dimensional diffraction grating pattern arranged with a predetermined pitch size;displacing by said deflection means, said scanning position in a second direction perpendicular to said first direction, and by an amount corresponding to said pitch size of said diffraction grating,scanning said electron beam in said first direction parallel to said one-dimensional diffraction grating pattern;detecting reflected electrons or secondary electrons emitted from said calibrating mark; andperforming, from said detection result, calibration for a deflection direction or a deflection quantity of said electron beam. 2. An electron-beam calibration method for an electron-beam system including at least an electron source for emitting an electron beam, a stage for mounting a sample thereon, deflection means, an objective lens, and a calibrating mark, said electron-beam calibration method comprising the steps of:scanning said electron beam on said calibrating mark in a first direction parallel to a one-dimensional diffraction grating pattern arranged with a predetermined pitch size;displacing said electron beam in a second direction perpendicular to said one-dimensional diffraction grating pattern,scanning said electron beam once again in first said direction parallel to said one-dimensional diffraction grating pattern;detecting reflected electrons or secondary electrons over an interval time, said reflected electrons or secondary electrons being emitted from said calibrating mark, said interval time corresponding to a time cycle for which said displacing coincides with a pitch size of said one dimensional diffraction grating pattern; andperforming, from said detection result, calibration for a deflection direction or a deflection quantity of said electron beam. 3. The electron-beam calibration method according to claim 1, further comprising the steps of:making a comparison between a reflected-electron image or a secondary-electron image acquired from said detection result and a reference image memorized in advance; andperforming, from said comparison result, said calibration for said deflection direction or said deflection quantity of said electron beam. 4. The electron-beam calibration method according to claim 3, wherein said detection result or said comparison result is a moiré pattern. 5. The electron-beam calibration method according to claim 1, wherein said pitch size of said one-dimensional diffraction grating pattern is a diffraction grating pattern pitch size determined based on an optical diffraction. 6. The electron-beam calibration method according to claim 1, wherein said one-dimensional diffraction grating pattern is of a superlattice multi-layer cross-section structure. 7. An electron-beam system, comprising:an electron source for emitting an electron beam;a deflector;an objective lens;a stage for mounting a sample thereon;a calibrating mark for calibrating an irradiation position with said electron beam; andan electron detector for detecting reflected electrons or secondary electrons generated by said irradiation with said electron beam, whereinsaid calibrating mark includes a one-dimensional diffraction grating pattern arranged with a predetermined pitch size,said electron-beam system further comprising:a control unit forscanning said electron beam in a first direction parallel to said one-dimensional diffraction grating pattern, anddisplacing said scanning position in a second direction and scanning said electron beam again in said first direction, said second direction being perpendicular to said one-dimensional diffraction grating pattern; anda display unit for displaying a reflected-electron image or a secondary-electron image acquired by said scanning. 8. The electron-beam system according to claim 7, whereinsaid control unit displaces said scanning position in said second direction and in correspondence with said pitch spacing of said diffraction grating. 9. The electron-beam system according to claim 7, whereinsaid control unit comprises:a signal analysis unit for making a comparison between said reflected-electron image or said secondary-electron image and a reference image memorized in advance, said reflected-electron image or said secondary-electron image being acquired by scanning said electron beam in said first direction parallel to said one-dimensional diffraction grating pattern; anda deflection control unit for correcting a deflection quantity of said electron beam based on said comparison result. 10. The electron-beam system according to claim 9, wherein said comparison result is displayed on said display unit. 11. The electron-beam system according to claim 7, whereinsaid control unit comprises:a signal processing unit for controlling said electron detector to detect said reflected electrons or secondary electrons with a period corresponding to said pitch spacing of said one-dimensional diffraction grating pattern, said reflected electrons or secondary electrons being emitted from said calibrating mark by said scanning of said electron beam;a signal calculation unit for calculating a scanning-position shift quantity from said reflected-electron image or said secondary-electron image detected and said one-dimensional diffraction grating pattern; anda deflection control unit for correcting a deflection quantity of said electron beam based on said calculation result. 12. The electron-beam system according to claim 7, wherein said pitch size of said one-dimensional diffraction grating pattern is a diffraction grating pattern pitch size determined based on an optical diffraction. 13. The electron-beam system according to claim 7, wherein said one-dimensional diffraction grating pattern is of a superlattice multi-layer cross-section structure. 14. The electron-beam system according to claim 7, wherein said calibrating mark is located on said stage.
	abstract	A new composition of matter includes 195mPt characterized by a specific activity of at least 30 mCi/mg Pt, generally made by method that includes the steps of: exposing 193Ir to a flux of neutrons sufficient to convert a portion of the 193Ir to 195mPt to form an irradiated material; dissolving the irradiated material to form an intermediate solution comprising Ir and Pt; and separating the Pt from the Ir by cation exchange chromatography to produce 195mPt.
062815083	abstract	A method and the associated apparatus for alignment and assembly of microlenses and microcolumns in which aligning structures such as rigid fibers are used to precisely align multiple microlens components. Alignment openings are formed in the microlens components and standard optical fibers are threaded through the openings in each microlens component as they are stacked. The fibers provide sufficient stiffness and stability to the structure to precisely align the apertures of the microlens components and thereby allow for increased assembly efficiency over traditional microlens and microcolumn bonding techniques.
	abstract	A seismic-resistant fuel storage system for a nuclear fuel pool includes a lined fuel pool and a fuel rack comprising tubular nuclear fuel storage cells attached to a common baseplate. Pedestals protrude downwardly from the baseplate supporting the rack on the pool base slab. Spaced embedment plates are fixedly anchored to the base slab to eliminate relative movement between the plates and pool liner. The embedment plates comprise upwardly open recessed receptacles each entrapping one of the rack pedestals therein. The receptacles are configured such that lateral movement of the fuel rack along the base slab in the event of a seismic event is constrained via engagement between receptacle walls and pedestals. Lateral seismic loads are not transferred to the pool liner. In some embodiments, the baseplates in the pool are coplanar and may be abutting engaged to mitigate rack movement during a seismic event.
	description	The present application is a continuation of and claims priority to U.S. Ser. No. 11/871,200 filed Oct. 12, 2007, which is a continuation of and claims priority to U.S. Ser. No. 11/164,121 that issued as U.S. Pat. No. 7,330,535 on Feb. 12, 2008, the disclosures of which are incorporated herein by reference. The present invention relates generally to radiographic imaging and, more particularly, to a beam chopper for a radiographic imaging system. The invention is also directed to an x-ray filter. The present invention is particularly related to photon counting and/or energy discriminating radiation detectors. Typically, in radiographic systems, an x-ray source emits x-rays toward a subject or object, such as a patient or a piece of luggage. Hereinafter, the terms “subject” and “object” may be interchangeably used to describe anything capable of being imaged. The x-ray beam, after being attenuated by the subject, impinges upon an array of radiation detectors. The intensity of the radiation beam received at the detector array is typically dependent upon the attenuation of the x-rays through the scanned object. Each detector element of the detector array produces a separate signal indicative of the attenuated beam received by each detector element. The signals are transmitted to a data processing system for analysis and further processing which ultimately produces an image. Generally, the x-ray source and the detector array are rotated about the gantry within an imaging plane and around the subject. X-ray sources typically include x-ray tubes, which emit the x-ray beam at a focal point. X-ray detectors typically include a collimator for collimating x-ray beams received at the detector, a scintillator for converting x-rays to light energy adjacent the collimator, and photodiodes for receiving the light energy from the adjacent scintillator and producing electrical signals therefrom. In a similar fashion, radiation detectors are employed in emission imaging systems such as used in nuclear medicine (NM) gamma cameras and Positron Emission Tomography (PET) systems. In these systems, the source of radiation is no longer an x-ray source, rather it is a radiopharmaceutical introduced into the body being examined. In these systems each detector of the array produces a signal in relation to the localized intensity of the radiopharmaceutical concentration in the object. Similar to conventional x-ray imaging, the strength of the emission signal is also attenuated by the inter-lying body parts. Each detector element of the detector array produces a separate signal indicative of the emitted beam received by each detector element. The signals are transmitted to a data processing system for analysis and further processing which ultimately produces an image. In most computed tomography (CT) imaging systems, the x-ray source and the detector array are rotated about a gantry encompassing an imaging volume around the subject. X-ray sources typically include x-ray tubes, which emit the x-rays as a fan or cone beam from the anode focal point. X-ray detector assemblies typically include a collimator for reducing scattered x-ray photons from reaching the detector, a scintillator adjacent to the collimator for converting x-rays to light energy, and a photodiode adjacent to the scintillator for receiving the light energy and producing electrical signals therefrom. Typically, each scintillator of a scintillator array converts x-rays to light energy. Each photodiode detects the light energy and generates a corresponding electrical signal. The outputs of the photodiodes are then transmitted to the data acquisition system and then to the processing system for image reconstruction. Conventional CT imaging systems utilize detectors that convert x-ray photon energy into current signals that are integrated over a time period, then measured and ultimately digitized. A drawback of such detectors is their inability to provide independent data or feedback as to the energy and incident flux rate of photons detected. That is, conventional CT detectors have a scintillator component and photodiode component wherein the scintillator component illuminates upon reception of x-ray photons and the photodiode detects illumination of the scintillator component, and provides an integrated electrical current signal as a function of the intensity and energy of incident x-ray photons. While it is generally recognized that CT imaging would not be a viable diagnostic imaging tool without the advancements achieved with conventional CT detector design, a drawback of these integrating detectors is their inability to provide energy discriminatory data or otherwise count the number and/or measure the energy of photons actually received by a given detector element or pixel. Accordingly, recent detector developments have included the design of an energy discriminating detector that can provide photon counting and/or energy discriminating feedback. In this regard, the detector can be caused to operate in an x-ray counting mode, an energy measurement mode of each x-ray event, or both. These energy discriminating detectors are capable of not only x-ray counting, but also providing a measurement of the energy level of each x-ray detected. While a number of materials may be used in the construction of an energy discriminating detector, including scintillators and photodiodes, direct conversion detectors having an x-ray photoconductor, such as amorphous selenium or cadmium zinc telluride, that directly convert x-ray photons into an electric charge have been shown to be among the preferred materials. A drawback of photon counting detectors, however, is that these types of detectors have limited count rates and have difficulty covering the broad dynamic ranges encompassing very high x-ray photon flux rates typically encountered with conventional CT systems. Generally, a CT detector dynamic range of 1,000,000 to one is required to adequately handle the possible variations in photon flux rates. In the very fast scanners now available, it is not uncommon to encounter x-ray flux rates of over 108 photons/mm2/sec when no object is in the scan field, with the same detection system needing to count only 10's of photons that manage to traverse the center of large objects. The very high x-ray photon flux rates ultimately lead to detector saturation. That is, these detectors typically saturate at relatively low x-ray flux levels. This saturation can occur at detector locations wherein small subject thickness is interposed between the detector and the radiographic energy source or x-ray tube. It has been shown that these saturated regions correspond to paths of low subject thickness near or outside the width of the subject projected onto the detector array. In many instances, the subject is more or less cylindrical in the effect on attenuation of the x-ray flux and subsequent incident intensity to the detector array. In this case, the saturated regions represent two disjointed regions at extremes of the detector array. In other less typical, but not rare instances, saturation occurs at other locations and in more than two disjointed regions of the detector. In the case of a cylindrical subject, the saturation at the edges of the array can be reduced by the imposition of a bowtie filter between the subject and the x-ray source. Typically, the filter is constructed to match the shape of the subject in such a way as to equalize total attenuation, filter and subject, across the detector array. The flux incident to the detector is then relatively uniform across the array and does not result in saturation. What can be problematic, however, is that the bowtie filter may not be optimum given that a subject population is significantly less than uniform and not exactly cylindrical in shape nor centrally located in the x-ray beam. In such cases, it is possible for one or more disjointed regions of saturation to occur or conversely to over-filter the x-ray flux and unnecessarily create regions of very low flux. Low x-ray flux in the projection results in a reduction in information content which will ultimately contribute to unwanted noise in the reconstructed image of the subject. Moreover, a system calibration method common to most CT systems involves measuring detector response with no subject whatsoever in the beam. This “air cal” reading from each detector element is used to normalize and correct the preprocessed data that is then used for CT image reconstruction. Even with ideal bowtie filters, high x-ray flux now in the central region of the detector array could lead to detector saturation during the system calibration phase. A number of imaging techniques have been proposed to address saturation of any part of the detector. These techniques include maintenance of low x-ray flux across the width of a detector array, for example, by modulating tube current or x-ray voltage during scanning However, this solution leads to increased scanned time. That is, there is a penalty that the acquisition time for the image is increased in proportion to the nominal flux needed to acquire a certain number of x-rays that meet image quality requirements. Other solutions include the implementation of over-range algorithms that may be used to generate replacement data for the saturated data. However, these algorithms may imperfectly replace the saturated data as well as contribute to the complexity of the CT system. It would therefore be desirable to design an x-ray flux management device that is effective in reducing detector saturation under high x-ray flux conditions while not compromising data acquisition under low x-ray flux conditions. The present invention is a directed an x-ray flux management device that overcomes the aforementioned drawbacks. The impact of radiographic detector design on radiographic image quality is foremost an issue of properly handling low-flux conditions (to effectively measure the limited x-ray transmission through thicker imaging regions). At the same time, the higher flux areas in these scans (such as detector readings through air and partially within the subject contours) also need to be correctly evaluated. If insufficient detector dynamic range is available, these high-flux detector channels tend to over-range and enter a saturated state. Since these over-range areas are typically in air or in highly irradiated portions of the subject, the exact measurement of each photon in these high-flux regions is not as critical as for the low-flux areas where each photon contributes an integral part to the total collected photon statistics and improved imaging quality. Subsequently, the invention addresses the specific needs of low- and high-flux regions and thereby provides the opportunity to use low dynamic range detectors for radiographic imaging. In this regard, the invention includes an x-ray flux management device that adaptively attenuates an x-ray beam to limit the incident flux reaching the subject and the radiographic detectors in the potentially high-flux areas while not affecting the incident flux and detector measurements in low-flux regions. While the invention is particularly well-suited for CT, the invention is also applicable with other x-ray imaging systems. In addition to reducing the required detector system dynamic range, the present invention provides an added advantage of reducing radiation dose. Therefore, in accordance with one aspect, the invention includes an x-ray beam chopper for a radiographic imaging apparatus. The beam chopper has a rotatable frame and at least one x-ray transmission window disposed in the rotatable frame that allows a generally free transmission of x-rays. The chopper also has at least one x-ray filtering window disposed in the rotatable frame that filters x-rays. In accordance with another aspect, the invention is directed to a radiographic imaging apparatus that includes an x-ray source and an x-ray detector. The apparatus further has a segmented filtering assembly having a generally annular frame with at least one low x-ray flux segment and at least one high x-ray flux segment, and a filtering assembly controller that causes the low x-ray flux segment to be in an x-ray beam path during a low x-ray flux data acquisition view and causes the high x-ray flux segment to be in the x-ray beam path during a high x-ray flux data acquisition view. According to another aspect, the invention includes an x-ray filter having a 3D semi-cylindrical rotatable filter body formed of x-ray attenuating matter. The filter also has a semi-conical bore formed in the 3D semi-cylindrical rotatable filter. The semi-conical bore has an elliptically shaped base. According to yet another aspect, the invention includes an x-ray filter assembly having a bowtie filter having an effective beam profile. The assembly further has a filter controller that tilts the bowtie filter during data acquisition to change the effective beam profile during data acquisition. Various other features and advantages of the present invention will be made apparent from the following detailed description and the drawings. The operating environment of the present invention is described with respect to a four-slice computed tomography (CT) system. However, it will be appreciated by those skilled in the art that the present invention is equally applicable for use with single-slice or other multi-slice configurations. Moreover, the present invention will be described with respect to the detection and conversion of x-rays. However, one skilled in the art will further appreciate that the present invention is equally applicable for the detection and conversion of other high frequency electromagnetic energy. Referring to FIGS. 1 and 2, an exemplary computed tomography (CT) imaging system 10 is shown as including a gantry 12 representative of a “third generation” CT scanner. Gantry 12 has an x-ray source 14 that projects a beam of x-rays 16 through an x-ray flux management assembly 17 toward a detector array 18 on the opposite side of the gantry 12. The x-ray flux management assembly will be described in greater detail with respect to FIGS. 3-12. Detector array 18 is formed by a plurality of detectors 20 which together sense the projected x-rays that pass through a medical patient 22. Each detector 20 produces an electrical signal that represents the intensity of an impinging x-ray beam and hence the attenuated beam as it passes through the patient 22. During a scan to acquire x-ray projection data, gantry 12 and the components mounted thereon rotate about a center of rotation 24. Rotation of gantry 12 and the operation of x-ray source 14 are governed by a control mechanism 26 of CT system 10. Control mechanism 26 includes an x-ray controller 28 that provides power and timing signals to an x-ray source 14 and a gantry motor controller 30 that controls the rotational speed and position of gantry 12. A data acquisition system (DAS) 32 in control mechanism 26 samples analog data from detectors 20 and converts the data to digital signals for subsequent processing. An image reconstructor 34 receives sampled and digitized x-ray data from DAS 32 and performs high speed reconstruction. The reconstructed image is applied as an input to a computer 36 which stores the image in a mass storage device 38. Computer 36 also receives commands and scanning parameters from an operator via console 40 that has a keyboard. An associated cathode ray tube display 42 allows the operator to observe the reconstructed image and other data from computer 36. The operator supplied commands and parameters are used by computer 36 to provide control signals and information to DAS 32, x-ray controller 28, gantry motor controller 30, and filter controller 31. In addition, computer 36 operates a table motor controller 44 which controls a motorized table 46 to position patient 22 and gantry 12. Particularly, table 46 moves portions of patient 22 through a gantry opening 48. The present invention is directed to an x-ray beam chopper that may be incorporated with the CT system described above or other radiographic systems, such as x-ray systems and the like. Generally, high-sensitivity photon counting radiation detectors are constructed to have a relatively low dynamic range. This is generally considered acceptable for proton counting detector applications since high flux conditions typically do not occur. In CT detector designs, low flux detector readings through the subject are typically accompanied by areas of high irradiation in air, and/or within the contours of the scan subject requiring CT detectors to have very large dynamic range responses. Moreover, the exact measurement of photons in these high-flux regions is less critical than that for low-flux areas where each photon contributes an integral part to the total collected photon statistics. Notwithstanding that the higher flux areas may be of less clinical or diagnostic value, images reconstructed with over-ranging or saturated detector channel data can be prone to artifacts. As such, the handling of high-flux conditions is also important. The present invention includes an x-ray flux management device designed to prevent saturation of photon counting x-ray systems having detector channels characterized by low dynamic range. Dynamic range of a detector channel defines the range of x-ray flux levels that the detector channel can handle to provide meaningful data at the low-flux end and not experience over-ranging or saturating at the high flux end. Notwithstanding the need to prevent over-ranging, to provide diagnostically valuable data, the handling of low-flux conditions, which commonly occur during imaging through thicker cross-sections and other areas of limited x-ray transmission, is also critical in detector design. As such, the x-ray flux management device described herein is designed to satisfy both high flux and low flux conditions. Referring now to FIG. 3, an x-ray flux management device according to one embodiment of the invention is shown. As illustrated, the device 17, which is shown relative to the z-axis or long axis of subject 22, is operative as an x-ray beam chopper that is positioned between x-ray tube 14 and z-plane collimator 50. In a preferred embodiment, the beam chopper 17 has a generally annular frame or tube 52 with two types of windows alternatively arranged along an outer rim thereof. In the illustrated exemplary embodiment, the generally annular frame is polygonal. One type of window is a transmission window 54 that provides unobstructed transmission of x-rays 16 and, as such, is designed to be placed in the x-ray beam path during low x-ray flux conditions, e.g. when a thicker subject cross-section is being imaged. The other window type is an x-ray filtering window 56 that filters or attenuates x-rays 16 when placed in the x-ray beam path and, as such, is designed to be placed in the x-ray beam path during high x-ray flux conditions, e.g. when a thinner subject cross-section is being imaged. In one embodiment, each x-ray filtering window 56 is composed of a block of x-ray filtering or attenuating material with holes (not shown) formed therein. The x-ray transmission windows 54 are preferably constructed to not effect the energy of the x-ray beam. In the exemplary embodiment of FIG. 3, the beam chopper has an octagonal frame. In this regard, the chopper is constructed to have four x-ray transmission windows 54 and four x-ray filtering windows 56. With this construction, the x-ray transmission windows 54 and x-ray filtering windows are alternately formed about the frame. As such, each x-ray transmission window is adjacent a pair of x-ray filtering windows. As further illustrated in FIG. 3, the transmission x-ray and x-ray filtering windows 54, 56 are arranged relative to or integrally formed within frame 52 such that the x-ray beam 16 passes through a pair of transmission windows 54 or a pair of filtering windows 56. With this orientation, transition times between adjacent windows are advantageously reduced. For example, for an octagonal beam chopper having four x-ray transmission windows and four x-ray filtering windows of substantially equal size, only a one-quarter rotation per data acquisition view is required. As such, a rotational speed of 30,000 rpm for one-half second scanners having 1,000 views per 360 degrees of acquisition is possible. As described above, the x-ray transmission windows 54 are placed in the x-ray beam path when the current data acquisition view is from a thicker subject cross-section. Conversely, the x-ray filtering windows 56 are placed in the x-ray beam path when the current data acquisition view is from a thinner subject cross-section. Accordingly, rotation of the chopper is dynamically controlled by controller 31, FIG. 2, to provide synchronization between chopper rotation and data acquisition. In this regard, it is contemplated that the chopper may be caused to rotate continuously at a fixed rotational speed or at a variable rotational speed. Additionally, it contemplated that the chopper may be initially held stationary with x-ray transmission windows placed in the x-ray beam. In this regard, saturation of the x-ray detector can be monitored and if the detector is at or near saturation, the chopper can be incrementally rotated such that x-ray filtering windows are placed in the x-ray path. For the next acquisition, the chopper is again rotated such that x-ray transmission windows are placed in the x-ray beam path. Saturation is again monitored and, if need be, a subsequent incremental rotation of the chopper. Accordingly, x-ray filtering windows are not placed in the x-ray beam path unless saturation is imminent or has occurred. Referring now to FIG. 4, position of the beam chopper 17 relative to the x-axis of subject 22 is illustrated. For purposes of simplicity, collimator 50, FIG. 3, is not shown. As illustrated, for the current data acquisition view, a pair of low x-ray flux or x-ray transmission windows 54 is positioned in the x-ray beam 16. At high x-ray flux conditions, the beam chopper 17 will be rotated by motor 58 to rotate x-ray filtering windows 56 into the x-ray beam path 16. In addition to rotating the beam chopper, it is contemplated that motor 58 may translate the beam chopper in the x-direction to accommodate asymmetrical subjects and variations in subject contours. In one preferred embodiment, motor 58 is a stepper motor. Referring now to FIG. 5, an alternate embodiment of beam chopper 17 is illustrated. In the illustrated embodiment, there are more x-ray transmission windows 54 than x-ray filtering windows 56. As shown, there is a 2:1 relationship between the number of x-ray transmission windows and the number of x-ray filtering windows. In this regard, only every third view would be attenuated if the beam chopper is continuously rotated. Accordingly, there is not an alternating between high x-ray flux views and low x-ray flux views as in the embodiment illustrated in FIG. 3. One skilled in the art will appreciate that such a 2:1 relationship between transmission and filtering views may be equivalently achieved with a chopper having equal number of transmission and filtering windows, but through variable rotational speed of the chopper such that the transmission windows are in the x-ray beam twice as long as the filtering windows. Also, it is contemplated that the beam chopper 17 may be constructed such that every Nth view is attenuated. In this regard, it is contemplated that the beam chopper can be designed to have NX transmission windows, where N is a number greater than one and X is the number of filtering windows. Referring now to FIG. 6, another embodiment of the beam chopper is illustrated. Similar to that illustrated in FIGS. 3 and 5, the beam chopper of FIG. 6 also has a generally annular frame 52 about which x-ray transmission windows 54 and x-ray filtering windows 56 are formed. Unlike the polygonal constructions previously described, the beam chopper 17 of FIG. 6 has a fixed radius. Notwithstanding this distinction, operation of the filter is similar to that previously described. The beam chopper 17 is rotated such that x-ray transmission windows 52 are in the x-ray beam path 16 during low x-ray flux conditions and x-ray filtering windows 54 are in the x-ray beam path 16 during high x-ray flux conditions. In the exemplary beam chopper illustrated in FIG. 6, there is a 2:1 relationship between transmission windows and filtering windows; however, it is contemplated that the beam chopper may have less than or more than a 2:1 ratio. As described above, it is contemplated that detector saturation readings may be acquired for a given view and if the detector has saturated (or will saturate), the beam chopper can be caused to rotate to place x-ray filtering windows in the x-ray beam. Thus, it is contemplated that for a saturated or near-saturated view, data may be acquired with the x-ray filtering windows in the x-ray beam path and that data can be used not only for image reconstruction but to correct the otherwise saturated data. Additionally, while the beam chopper has been described such that either two x-ray transmission windows or two x-ray filtering windows are in the x-ray beam at any given moment, it is contemplated that the beam chopper may be constructed such that only one transmission or only one filtering window is in the beam path. That is, it is contemplated that the windows may be formed on a hemispherical frame such that through pendulum-like translation, different attenuation profiles may be presented. In this regard, it is further contemplated that more than two types of windows may be supported by the frame. The invention contemplates that various windows of different attenuation power may be supported by the frame whereby the continuum of attenuation windows ranges from a free transmission window of zero attenuation to a maximum attenuation window. Moreover, it is contemplated that such a hemispherical frame could be caused to rotate clockwise as well as counter-clockwise and at a fixed or variable speed. Additionally, it is contemplated that a mechanical shutter of x-ray filtering material may be dynamically presented in the x-ray beam during high x-ray flux conditions. The present invention also includes an inventive bowtie filter. Standard bowtie filters have a symmetrical, one-dimensional shape. To overcome limitations associated with these standard bowtie filters, the present invention is also directed to a 3D semi-cylindrical rotatable bowtie filter. This multi-dimensional filter 60, shown in FIG. 7, has a cylindrical frame 62 with a semi-conic bore 64 formed therein. The bore 64 has an elliptical base 66. This is in stark contrast to conventional bowtie filters which have a circular base. Additionally, also in contrast to conventional bowtie filters, filter 60 is not symmetrical. This is illustrated by the cross-sectional views of FIGS. 8 and 9. Referring now to FIG. 8, cross-sectional views of filter 60 taken along lines 8-8 and lines 9-9, respectively, are shown. As illustrated, filter 60 is constructed to have a bore 64 formed within frame 62. The width of the bore 64 cut along line 8-8, however, is greater than that of bore cut along line 9-9. This results in a different absorption profile for any rotational angle of the filter 60. Also, it is contemplated that the filter may be dynamically repositioned during data acquisition so that the resulting profile can be matched to the subject's body and, in particular, centered for non-centered subjects. In this regard, it is contemplated that precise positioning of the subject can be measured and used to control translation of the filter. Precise positioning can be determined from positioning sensors, scout scan data, and the like. By doing so, the present invention supports rotation and translation of the filter during data acquisition to track subject profile. It is also contemplated that multiple filters in a stacked arrangement could be used and moved in tandem or independently to achieve a desired attenuation profile. This can be particularly advantageous when imaging two legs and other anatomical structures that require a relatively complex attenuation profile. Referring now to FIGS. 10-11, a filter assembly in accordance with another embodiment of the present invention is shown. In this embodiment, a pair of bowtie filters 68, 70 are shown relative to the x-axis and in x-ray beam 16. Each filter 68, 70 is thicker in the z-direction than conventional bowtie filters. In contrast to conventional bowtie filters, however, filter 68, 70 are designed to be tilted by a tilt mechanism (not shown) to effectively change the attenuation profile of the filters. In addition to being tilted, the filters may also be moved laterally in the x-direction to better match a given subject's contours or accommodate a non-centered subject. Additionally, while two filters stacked on top of another are shown, it is contemplated that less than two or more than two filters may be used. As illustrated in FIG. 11, filters 68, 70 are tiltable relative to the z-axis. In this regard, the attenuation profile generated by the filters 68, 70 can be dynamically controlled to match a desired attenuation profile. The tilt angle (and translation) position of the bowtie filters can be changed during data acquisition to track a given subject profile. In a preferred embodiment, the filters can be tilted a maximum ninety degrees. This ninety degree tilt range defines a minimum absorption profile at zero degrees to a maximum absorption profile at ninety degrees. Referring now to FIG. 12, package/baggage inspection system 72 includes a rotatable gantry 74 having an opening 76 therein through which packages or pieces of baggage may pass. The rotatable gantry 74 houses a high frequency electromagnetic energy source 78 as well as a detector assembly 80. A conveyor system 82 is also provided and includes a conveyor belt 84 supported by structure 86 to automatically and continuously pass packages or baggage pieces 88 through opening 76 to be scanned. Objects 88 are fed through opening 76 by conveyor belt 84, imaging data is then acquired, and the conveyor belt 84 removes the packages 88 from opening 76 in a controlled and continuous manner. As a result, postal inspectors, baggage handlers, and other security personnel may non-invasively inspect the contents of packages 88 for explosives, knives, guns, contraband, etc. Therefore, in accordance with one embodiment, the invention includes an x-ray beam chopper for a radiographic imaging apparatus. The beam chopper has a rotatable frame and at least one x-ray transmission window disposed in the rotatable frame that allows a generally free transmission of x-rays. The chopper also has at least one x-ray filtering window disposed in the rotatable frame that filters x-rays. In accordance with another embodiment, the invention is directed to a radiographic imaging apparatus that includes an x-ray source and an x-ray detector. The apparatus further has a segmented filtering assembly having a generally annular frame with at least one low x-ray flux segment and at least one high x-ray flux segment, and a filtering assembly controller that causes the low x-ray flux segment to be in an x-ray beam path during a low x-ray flux data acquisition view and causes the high x-ray flux segment to be in the x-ray beam path during a high x-ray flux data acquisition view. According to another embodiment, the invention includes an x-ray filter having a 3D semi-cylindrical rotatable filter body formed of x-ray attenuating matter. The filter also has a semi-conical bore formed in the 3D semi-cylindrical rotatable filter. The semi-conical bore has an elliptically shaped base. According to yet another embodiment, the invention includes an x-ray filter assembly having a bowtie filter having an effective beam profile. The assembly further has a filter controller that tilts the bowtie filter during data acquisition to change the effective beam profile during data acquisition. While the present invention is applicable with a number of radiographic imaging systems, it is particularly well-suited for CT systems and, especially, those systems having detectors with relative small dynamic range, such as photon counting and energy discriminating detectors. In this regard, the present invention is believed to be a key enabler for the use of direct conversion and energy discriminating/photon counting detectors with conventional CT systems. Additionally, as the beam chopper and filters selectively limit radiation exposure, the invention advantageously reduces subject dose without sacrificing image quality. The present invention has been described in terms of the preferred embodiment, and it is recognized that equivalents, alternatives, and modifications, aside from those expressly stated, are possible and within the scope of the appending claims.
	claims	1. A radiation collimator arranged between a radiation source and an object, comprising:a plurality of absorber channels that are arranged adjacent to one another;a plurality of first absorber elements arranged in the plurality of absorber channels that blocks a radiation emitted by the radiation source in a first position and passes the radiation through the absorber channels in a second position; anda two-dimensional collimator aperture that is formed by the plurality of first absorber elements arranged in the plurality of absorber channels;wherein the plurality of first absorber elements are each configured to be rod-shaped and to be moved in a respective absorber channel from the first position into the second position by a rotation of the respective first absorber element about a rotation axis aligned parallel to a longitudinal axis of the respective absorber channel;and wherein the plurality of first absorber elements are respectively shaped as a cone or a truncated pyramid. 2. The radiation collimator as claimed in claim 1, wherein the radiation reaches the object through the plurality of absorber channels by moving the plurality of first absorber elements in a radiation direction. 3. The radiation collimator as claimed in claim 1, further comprising a plurality of second absorber elements, wherein a respective second absorber element is fixedly arranged in the respective absorber channel downstream or upstream of the respective first absorber element. 4. The radiation collimator as claimed in claim 3, wherein the respective first absorber element and the respective second absorber element are displaceable arranged in respect of one another. 5. The radiation collimator as claimed in claim 3, wherein the respective first absorber element and the respective second absorber element are rotatable arranged in respect of one another. 6. The radiation collimator as claimed in claim 3, wherein the respective absorber channel and the respective first absorber element and the respective second absorber element taper in a direction of the radiation source. 7. The radiation collimator as claimed in claim 1, further comprising a plurality of second absorber elements, wherein a position of the respective first absorber element is changeable with a position of the respective second absorber element. 8. The radiation collimator as claimed in claim 1, wherein the collimator aperture has a partial surface of a surface of a sphere. 9. The radiation collimator as claimed in claim 8, wherein a central point of the collimator aperture is a focal point of the radiation source. 10. The radiation collimator as claimed in claim 1, wherein the radiation collimator is an x-ray collimator and the radiation is an x-ray radiation. 11. The radiation collimator as claimed in claim 1, wherein the radiation collimator is used in a radiation therapy apparatus.
041464290	description	DETAILED DESCRIPTION OF THE SHOWN EMBODIMENT Referring particularly to the drawings, there is shown in FIG. 1 the building foundation 10 of a nuclear reactor (not shown). Schematically, the nuclear reactor fissionable core is shown as mass 12, actually the mass 12 would be located at some position above the foundation 10. The upper end of the foundation 10 and located under the mass 12 is a receiving funnel assembly 14. The funnel assembly 14 can take the form of a single enlarged funnel or take the form of a plurality of separate funnels, depending upon the particular installation. The outlet 16 of the funnel assembly 14 leads to a plurality of first passageways 18, 20 and 22. The total cross-sectional area of the passageways 18, 20 and 22 is to be approximately equal to the cross-sectional area of the passageway 16. Although three in number of first passageways are shown, it is considered to be within the scope of this invention that more or less than three can be employed. In the area of the junction of the first passageways 18, 20 and 22 with the passageway 16 there is located a quantity 24 of lead or tin or alloys thereof. It is also considered to be within the scope of this invention that other easily meltable materials could be used. Again, it is to be reiterated that the function of the material 24 is to facilitate even distribution of the molten fissionable material which is moved through passageway 16 and also, the material 24 is to intermix with the fissionable material to assist in decreasing the reaction. Also, the use of the meltable material functions to increase the fluidity of the fissionable material so it will continue to flow and distribute through the passages until it is assured that a safe, low temperature for the fissionable material has been achieved. Although the meltable material will normally be located only at the junctions of the passageways, it may also be located throughout the passageways. If the easily meltable material (such as oxides) was such that it was displaced and not mixable with the liquid fissionable material, the meltable material would rise to the upper surface of the liquid fissionable material. As the fissionable material continued to move downwardly, the oxide material would provide a radiation shield preventing escape of radiation contaminates through the upper passages and hence to the ambient. The radiation shield may occur by the oxide material rehardening and assuming a particular location or by remaining liquid and following the fissionable material in its downward path. It is to be apparent that the subject matter of this invention will only be used in an emergency situation in which the solid mass of fissionable material 12 is caused to go out of control and become molten and begin to move under the effect of gravity. The first passageway 18 terminates into a plurality of second passageways 26, 28 and 30. The first passageways 20 and 22 also lead to separate second passageway structure. Again, the number of second passageways is considered to be a matter of choice. Also, in normal practice, the totally combined cross-sectional area of the second passageways will be approximately equal to the cross-sectional area of its connected first passage. There will also be a quantity of easily meltable material 32 placed within the area of the junction of the passageways 26, 28 and 30 with the first passageway 18. The same is true also for the junction of the second passages with the first passages 20 and 22. This passageway configuration is continued in a branch manner with the second passageway 26 being connected to a plurality of third passages 34, 36 and 38. The same procedure is true for each of the second passageways, with there also being a similar manner an easily meltable material 40 located in the area of the junction of the third passages with the second passage. In actual practice, it may be sufficient to end with third passages which are connected to second passageway 28. However, in other instances it may be desirable to have each third passageway connect to a plurality of fourth passageways, such as the connection of passages 42, 44 and 46 with passageway 34. The previously mentioned cross-sectional area arrangement is to be normally maintained at the junction of each set of passages. Also, at the junction of each set of passages there may also be included easily meltable material. In normal practice, it is envisioned that upon the material achieving the dispersement within the separated passages of the third or fourth passages, that sufficient dispersement is achieved so that the reaction has been brought into control and the heat of reaction will be lowered sufficiently so that the cement material of the foundation will not be caused to melt and will function as a plurality of separate spaced apart containers with a quantity of fissionable material. However, as an option, it may be desirable to include at the end of each fourth passage (at the end of the third passage if there are no fourth passages used) an explosive device 48. The explosive device 48 will normally take the form of a conventionally available explosive apparatus and the details of which are not believed necessary to discuss in relation to the structure of this invention. The explosive will normally be designed to be activated upon the fissionable material coming into contact therewith which will be designed to very finely distribute the fissionable material within a base layer 50 of loosely packed material, such as sand. Such a complete dispersement of the fissionable material would insure that the reaction would be completely under control.
	claims	1. A method of generating a medical isotope comprising:providing an annular reaction vessel holding an aqueous fissile solution, the annular reaction vessel having an inner cylindrical wall and an outer cylindrical wall, and a particle emitter positioned inside a cylindrical space defined by the inner cylindrical wall;directing a beam of charged particles along a central axis of the annular reaction vessel into the particle emitter positioned concentrically along the central axis within the annular reaction vessel to produce a beam of neutrons passing radially outward from the particle emitter through the annular reaction vessel into the aqueous fissile solution to produce a nuclear fission reaction dominated by neutrons bombarding the aqueous fissile solution and producing with the neutrons a fission reaction in the aqueous fissile solution to produce additional neutrons in the annular reaction vessel; andcooling the inner and outer cylindrical walls of the annular reaction vessel to provide cooling of the aqueous fissile solution during the nuclear fission reaction within the annular reaction vessel by a cooling fluid. 2. The method of claim 1 wherein the medical isotope is 99Mo and the method further comprises extracting the 99Mo from the aqueous fissile solution. 3. The method of claim 1 wherein cooling of the inner and outer cylindrical walls of the annular reaction vessel is by circulating the cooling fluid through the first and second annular cooling jackets, the first annular cooling jacket abutting and surrounding the outer cylindrical wall of the annular reaction vessel and the second annular cooling jacket abutting the inner cylindrical wall of the annular reaction vessel. 4. The method of claim 3 further comprising providing a feedback controller comprising at least one temperature probe in thermal communication with at least one of the first and second annular cooling jackets and a valve in fluid communication with the aqueous fissile solution and actuating the valve to adjust a level of the aqueous fissile solution based on a temperature sensed by the temperature probes. 5. The method of claim 1 wherein the aqueous fissile solution contains low enriched uranium. 6. The method of claim 5 wherein the low enriched uranium has a concentration between 10 and 450 grams of low enriched uranium per liter solution. 7. The method of claim 5 wherein the aqueous fissile solution contains a mixture of water and at least one of uranyl nitrate, uranyl sulfate, uranyl fluoride or uranyl phosphate. 8. The method of claim 1 further comprising a neutron multiplier absorbing neutrons from the particle emitter traveling outward and releasing more neutrons than the neutron multiplier absorbs. 9. The method of claim 8 wherein the neutron multiplier is selected from a group consisting of beryllium, depleted uranium, and natural uranium. 10. The method of claim 9 wherein the neutron multiplier provides for a neutron multiplication of 1.5-3.0. 11. The method of claim 1 further comprising a neutron moderator absorbing neutrons traveling inward toward the particle emitter and reducing a speed of the neutrons that the neutron moderator absorbs. 12. The method of claim 1 further comprising producing the beam of charged particles by ionizing a gas by at least one of a microwave emission, ion impact ionization, and laser ionization. 13. The method of claim 12 wherein the beam of charged particles directed into the particle emitter are deuterium ions (D+). 14. The method of claim 1 further comprising an accelerator receiving the beam of charged particles and accelerating the beam of charged particles toward the particle emitter. 15. The method of claim 1 further comprising reflecting neutrons back into the annular reaction vessel. 16. The method of claim 1 wherein the cooling fluid is chilled water. 17. The method of claim 1 wherein the particle emitter is a target material receiving the beam of charged particles from a particle source. 18. The method of claim 1 wherein an aspect ratio defined by a radial thickness of the annular reaction vessel perpendicular to the axis to a height of the annular reaction vessel along the axis is greater than 0.1. 19. The method of claim 18 wherein the aspect ratio is between 0.1 and 0.3. 20. The method of claim 19 wherein the aspect ratio is between 0.12 and 0.25.
046577297	abstract	A solid tag material which generates stable detectable, identifiable, and measurable isotopic gases on exposure to a neutron flux to be placed in a nuclear reactor component, particularly a fuel element, in order to identify the reactor component in event of its failure. Several tag materials consisting of salts which generate a multiplicity of gaseous isotopes in predetermined ratios are used to identify different reactor components.
	claims	1. A field reversed magnetic field configuration fusion reactor system comprising:a reactor chamber in a field reversed magnetic field configuration;a gas injection system for injecting deuterium and helium-3 fuel into the reactor chamber for fusion reactions;a plurality of radio frequency (RF) antennae configured to generate an odd-parity rotating magnetic field capable of causing a plasma to heat to a temperature sufficient to cause fuel ions to fuse to produce fusion products;a plurality of superconducting flux coils around the reactor chamber in which an induced current is generated in response to the odd-parity rotating magnetic field, wherein the induced current generates a magnetic confinement field that magnetically confines the plasma; anda direct energy conversion system that is configured to extract energy from the fusion products, resulting from fusion reactions in the plasma, that pass through a scrape-off layer. 2. The system of claim 1, wherein power is extracted from Bremsstrahlung radiation and synchrotron radiation generated by the fusion reactions using a heat engine and high temperature heat exchangers. 3. The system of claim 1, wherein the scrape-off layer is configured to allow the fusion products resulting from fusion reactions to pass through before the energy from the fusion products is extracted in the direct energy conversion system.
	description	This application claims the benefit of U.S. provisional application Ser. No. 60/470,316 filed May 14, 2003, which is incorporated herein by reference. The invention relates to the diagnostic imaging arts. It finds particular application in conjunction with defining measurement periods for data intervals in CT scanners and will be described with particular reference thereto. However, it is to be appreciated that the invention is also amenable to other applications. Analog/digital (A/D) conversion in a CT scanner utilizes an integrating current to frequency converter (IFC). The IFC is a current-controlled oscillator. The current produced by a detector associated with the CT scanner varies the frequency of the current-controlled oscillator. During a data interval (which is defined by the angular position of a rotating gantry or, more precisely, an arc segment), the IFC pulses are counted, and the time from the first pulse to the last pulse is measured with high precision. The actual measurement is calculated by taking the ratio of the COUNTS to the TIME. The precision of the measurement is high since it is determined by the precision of the TIME count which is produced by counting the pulses from a high frequency oscillator. In a “delta data” mode of operation, the counting of COUNTS and TIME pulses starts with the last IFC pulse of the preceding data interval and ends with the last IFC pulse of the measured data interval. By allowing the measurement period to extend into the preceding data interval, all the current from the radiation detector is utilized, thus insuring high quantum signal to noise ratio. The “delta data” technique does, however, advance (or skew) the measurement period from its physical arc segment (i.e., data interval). With a large number of COUNT pulses in the data interval, this shift is minimal. If 100 COUNT pulses are counted, the skewing is nominally 0.5%. However, for low signal levels, this skew can be significant. If only one pulse is generated per data interval, the skewing is nominally 50% but can be up to 100%. This data skewing may cause objectionable image artifacts. A standard ratiometric type A/D conversion (without delta data) requires that at least two COUNT pulses be produced per data interval. When employing delta data this requirement is reduced to one COUNT pulse per data interval. In order to insure that the minimum pulse rate is maintained, an offset dc current is injected into the front end. The counts resulting from this offset current are subsequently subtracted out before taking the ratio of COUNT to TIME. However, the shot noise associated with this offset current increases the input noise of the A/D conversion thus reducing the overall dynamic range of the system. There is, therefore, a need to improve the accuracy of previous delta data modes by reducing (or eliminating), on the average, the skewing of the measurement period with respect to the measured data interval. There is also a desire to further reduce the required offset current in order to minimize noise and improve the dynamic range of the system. In one embodiment of the invention, a CT scanner includes a means for rotating a radiation source around an examination region, a means for generating an analog data signal that varies with an intensity of radiation traversing the examination region, a means for converting the analog data signal to a digital data signal including aperiodic pulses varying in frequency with the intensity of the radiation traversing the examination region as the radiation source rotates about the examination region, a means for producing a time signal indicative of data intervals, and a means for determining average radiation intensity in each data interval by counting the pulses of the digital data signal starting with a digital data signal pulse occurring in a preceding data interval and continuing to a digital data signal pulse occurring in a succeeding data interval. In another embodiment, the invention provides a method of measuring an intensity of detected radiation in a CT scanner. A radiation source is rotated around an examination region. An analog data signal that varies with an intensity of radiation traversing the examination region is generated. The analog data signal is converted to a digital data signal including aperiodic pulses varying in frequency with the intensity of the radiation traversing the examination region as the radiation source rotates about the examination region. A time signal indicative of data intervals is produced. Average radiation intensity in each data interval is determined by counting the pulses of the digital data signal starting with a digital data signal pulse occurring in a preceding data interval and continuing to a digital data signal pulse occurring in a succeeding data interval. In still another embodiment of the invention, an apparatus for measuring an intensity of a detected radiation in a CT scanner includes a channel circuit which generates time-based digital information from an analog data signal for a measured data interval, the time-based digital information including at least one component of the analog data signal from a preceding data interval and a succeeding data interval, a storage circuit which stores the time-based digital information, a control circuit which determines when to store the time-based digital information, and a processor which determines an average intensity of the detected radiation for the measured data interval from the stored time-based digital information. One advantage of the invention is the measurement period for a measured data interval is, on the average, centered on the data interval, thus producing an average measurement skew of zero. Still another advantage is, under conditions of high attenuation, the measurement period is significantly longer than the data interval thus producing more integrated signal, reducing quantum noise, and increasing the system dynamic range. Yet another advantage is the increase in measurement period as the input signal decreases produces an adaptive filtering effect in the analog domain that can potentially improve image quality by reducing noise more effectively than by subsequently filtering in the digital domain. Still yet another advantage is, in various embodiments, offset current can be reduced to a point where less than one pulse occurs per data interval. This reduces shot noise associated with the offset current and decreases the effects of quantization noise and 1/f noise. The resulting overall noise reduction improves image quality and extends the dynamic range of the system. Other advantages will become apparent to those of ordinary skill in the art upon reading and understanding the following detailed description. With reference to FIG. 1, a CT scanner 10 includes a stationary gantry 12 a rotating gantry 14, an imaging region 16, a radiation source 20, a collimator and shutter assembly 22, a subject support 30, a head restraint 32, a plurality of radiation detectors 40 or 42, an encoder 44, a signal processor 46, a reconstruction processor 48, a volume image memory 50, a video processor 52, and a display device 54. The stationary gantry 12 and rotating gantry 14 define the imaging region 16. The rotating gantry 14 is supported by the stationary gantry 12 for rotation about the examination region 16. The radiation source 20 (e.g., x-ray tube) is arranged on the rotating gantry 14 for rotation therewith. The radiation source 20 produces a beam of penetrating radiation that spans and passes through the examination region 16 as the rotating gantry 14 is rotated by an external motor (not illustrated) about a longitudinal axis of the examination region 16. The collimator and shutter assembly 22 forms the beam of penetrating radiation into a fan, cone, or wedge shape and selectively gates the beam on and off. Alternately, the radiation beam is gated on and off electronically at the radiation source 20. The patient support 30, such as a radiolucent couch or the like, suspends or otherwise holds a subject being examined or imaged at least partially within the examination region 16 such that the beam of radiation defines a volume through the region of interest of the subject. The head restraint 32 restricts the mobility of the subject's head. In a third generation CT scanner, an arc or a 2-dimensional array of radiation detectors 40 is mounted peripherally across from the radiation source 20 on the rotating gantry 14. In a fourth generation CT scanner, one or more stationary rings of radiation detectors 42 are mounted around the stationary gantry 12. Regardless of the configuration, the radiation detectors 40, 42 are arranged to receive the radiation emitted from the radiation source 20 after it has traversed the imaging region 16. The radiation detectors 40, 42 convert the detected radiation into analog data signals. That is, each radiation detector 40, 42 produces an analog data signal that is proportional to an intensity of received radiation. The signal processor 46 receives the analog data signals from the radiation detectors 40, 42. The signal processor 46 optionally performs filtering and other operations (e.g., generation of time-based digital information and calculation of average radiation intensity per data interval) before passing the data to a reconstruction processor 48 that reconstructs volume image representations of the subject for storage in a volume image memory 50. A video processor 52 under operator control retrieves and formats selected portions of the data for display on a display device 54, printing on a printer, etc. as a slice image, 3-dimensional rendering, or the like. During each orbit of the rotating gantry 14, the encoder 44 produces an index signal that is transmitted to the signal processor 46 to associate the position or angular arc segments of the rotating gantry with the analog data signals from the radiation detectors 40, 42. Each rotation of the radiation source is broken up into a succession of individual scan segments (i.e., data intervals) as the rotating gantry 14 turns or orbits the subject. In the preferred embodiment the index signal is a series of pulses, with a predetermined amount of pulses for each data interval. The last pulse for each data interval indicating termination of one data interval and initiation of a next or succeeding data interval. In alternate embodiments, devices capable of producing a similar index signal may be used in place of the encoder 44. The encoder 44 produces an index signal pulse at regular angular intervals, e.g. 0.1 degree. The index signal provides a timing signal defining the beginning and end of successive data intervals. The signal processor 46 includes a plurality of delta data channel circuits 56a-56n that are each responsive to individual analog data signals from the radiation detectors 40, 42, a delta data control circuit 58 that is responsive to an index signal from the encoder 44, a delta data storage circuit 60 for accumulating time-based digital information corresponding to the analog data signals, a delta data processor 62, and a radiation intensity storage circuit 64. The delta data channel circuits 56a-56n are typically identically constructed. With reference to FIG. 2, only one delta data channel circuit 56a is shown in the signal processor 46 to simplify the description. The delta data control circuit 58 develops time-based digital information in the delta data channel circuit 56a corresponding to the analog data signal. This type of conversion may also be referred to as a analog-to-digital (A/D) conversion. In principle, this type of conversion is preferably accomplished using current to frequency conversion (IFC) or voltage to frequency conversion (VFC) techniques. At appropriate times, the delta data control circuit 58 transfers the time-based digital data information from the delta data channel circuit 56a to the delta data storage circuit 60 and notifies the delta data processor 62 when the data is ready to be read from the delta data storage circuit 60. The delta data channel circuit 56a includes a summing module 66, an offset module 68, an IFC 70, a data pulse detector 72, a free-running oscillator 74, a data counter 76, and a time counter 78. The delta data control circuit 58 includes a data interval detector 80 and a delta data controller 82. In summary, the delta data channel circuit 56a provides an A/D conversion of the analog data signals by integrating a current output and producing a pulse train of a corresponding frequency. The delta data control circuit 58 monitors the data interval index signal and the output of the IFC 70. During scanning operations, the intensity of the analog data signal inherently varies with tissue density. In one embodiment, the delta data control circuit 58 stores “start data count” and “start time” for each data interval in response to the last “pulse” of the pulse data signal in the preceding data interval. In this embodiment, the delta data control circuit 58 also stores the “end data count” and “end time” for each data interval in response to the first pulse of the succeeding data interval. The delta data processor 62 determines the number of pulses (i.e., COUNTS) from the difference between the “end data count” and the “start data count” and the difference between “end time” and “start time” (i.e., TIME) for each data interval. The delta data processor 62 divides the COUNTS by the TIME to generate a numeric radiation intensity value for one data interval for one detector. Thus, the measurement period reflected by the COUNTS and TIME for each data interval extends into both the preceding and succeeding data intervals. This is referred to as a symmetrical delta data mode of operation. Although the measurement period extends outside of the measured data interval, on the average, the measurement periods are centered on the measured data intervals. In this manner, the sampling window dynamically widens beyond one data interval with high attenuation. The longer measurement periods reduce noise and increase the S/N ratio during high attenuation (when a higher S/N is most important), thus increasing the overall dynamic range of the A/D conversion. This technique produces a symmetrical variable filtering method for measuring radiation intensity during scanning operations. In the embodiment being described, the delta data channel circuit 56a preferably guarantees that at least one “pulse” is output from the IFC 70 during each data interval. To accomplish this, the offset module 68 provides an offset current to the summing module 66. The summing module 66 combines the offset current with the analog data signal to produce an offset data signal. Preferably, the current provided by the offset module 68 is adjusted to a minimum level required to guarantee that the IFC 70 generates at least one pulse during each data interval. The data pulse detector 72 monitors the output of the IFC to detect pulses. Each time a pulse is detected, the detected event is communicated to the delta data controller 82. The IFC 70 provides a digital pulse train output (i.e., pulse data signal) that varies in frequency based on the level of the offset data signal. As such, the pulse data signal is a digital representation of the analog data signal. The pulse data signal is provided to the data counter 76. The data counter 76 counts each pulse and accumulates a “data count.” In an alternate embodiment, the “data count” can be based on voltage rather than current. In this alternate embodiment, the IFC is replaced with a VFC. In the embodiment being described, the oscillator 74 of the delta data channel circuit 56a is free-running and provides a digital pulse train (i.e., time signal) at a relatively constant high frequency to the time counter 78. As such, the time signal is a digital representation of elapsed time. The time counter accumulates a “time count.” The combination of the “data count” and the “time count” provides time-based digital information representative of the radiation passing through a subject during a scanning operation. In an alternate embodiment, the oscillator 74 and time counter 78 may be combined as a time circuit, separate from the delta data channel circuits 56a-56n, that is common to each delta data channel circuit. In the embodiment being described, the data interval detector 80 receives the index signal from the encoder 44 and detects the rising edge generated during movement of the rotating gantry 14. Each pulse indicates the end of one data interval and the start of the next data interval. Each time the rising edge of a pulse is detected, the event is communicated to the delta data controller 82. The delta data controller 82 uses the combination of events detected by the data pulse detector 72 and data interval detector 80 to determine when to process the contents of the data counter 76 and the time counter 78 with the delta data processor 62 to develop the intensity value for each data interval. Since the time-based digital information developed by the signal processor 46 includes data from preceding and succeeding data intervals for a measured data interval, the delta data controller 82 and the delta data processor 62 may process information associated with three consecutive data intervals at any given time. The following description discusses how information for the three consecutive data intervals is processed by referencing the second, third, and fourth data intervals respectively. Information associated with the second data interval actually starts during a first data interval. The delta data controller 82 communicates a “store” signal to the data counter 76 and time counter 78 each time a “pulse” is detected by the data pulse detector 72. The “store” signal directs the data counter 76 and time counter 78 to transfer their current values (i.e., “data count” and “time count”) to the delta data storage circuit 60. The delta data controller 82 also communicates address information associated with the delta data storage circuit 60 identifying locations in the delta data storage circuit 60 where the data counter 76 and time counter 78 are to store their current values. During the first data interval, the address information identifies storage locations for the “start data count” and the “start time” for the second data interval. In response to the store signal and address information, the data counter 76 stores its current value in the “start data count” location for the second data interval and the time counter 78 stores its current value in the “start time” location for the second data interval. If a subsequent “pulse” on the pulse data signal is detected before the next index pulse is detected by the data interval detector 80, the “start data count” and “start time” locations for the second data interval are overwritten in the same manner. When the next index pulse is detected by the data interval detector 80, the rotating gantry 14 has reached the second data interval and the address information in the delta data controller 82 is altered to identify storage locations for the “start data count” and the “start time” for the third data interval. During the second data interval, each time the start of a “pulse” on the pulse data signal is detected by the data pulse detector 72, the delta data controller 82 communicates the “store” signal and associated address information to the data counter 76 and time counter 78 in the same manner as describe above. However, the data counter 76 stores its current value in the “start data count” location for the third data interval rather than overwriting the value stored for the second data interval. Likewise, the time counter 78 stores its current value in the “start time” location for the third data interval rather than overwriting the value stored for the first data interval. The “start data count” and “start time” locations for the third data interval are overwritten in the same manner if a subsequent “pulse” on the pulse data signal is detected before the next index pulse is detected by the data interval detector 80. When the next index pulse is detected by the data interval detector 80, the rotating gantry 14 has reached the third data interval and the address information is altered to identify storage locations for the “end data count” and “end time” for the second data interval and the “start data count” and “start time” for the fourth data interval. When the start of a first “pulse” on the pulse data signal is detected during the third data interval, the delta data controller 82 communicates the “store” signal and associated address information to the data counter 76 and time counter 78 in the same manner as described above. However, the data counter 76 stores its current value in both the “end data count” location for the second data interval and the “start data count” location for the fourth data interval. Likewise, the time counter 78 stores its current value in both the “end time” location for the second data interval and the “start time” location for the fourth data interval. The “start data count,” “end data count,” “start time,” and “end time” for the second data interval are now stored in the delta data storage circuit 60. At this point, the delta data controller 82 communicate a read signal and associated address information to the delta data processor 62. The read signal indicates that the stored “start data count,” “start time,” “end data count,” and “end time” for the second data interval are ready to be read from the delta data storage circuit 60. The address information identifies the “start data count,” “start time,” “end data count,” and “end time” locations from which to read the time-based digital information for the second data interval. The delta data processor 62 subtracts the “start data count” from the “end data count” to determine the COUNT for the second data interval and subtracts the “start time” from the “end time” to determine the TIME for the second data interval. These values for COUNT and TIME relate to an average level of intensity for the combined offset current and analog data signals during the second data interval. The counts produced by the offset current are subtracted from the COUNT, and the result is divided by the TIME to determine the intensity of the detected radiation for the second data interval. The radiation intensity values for each detector and each data interval are stored in the radiation intensity storage circuit 64 awaiting reconstruction by the reconstruction processor 48. At this point, the delta data processor 62 may communicate a read signal and associated address information to the reconstruction processor 48. The read signal indicates that the stored radiation intensity value for the second data interval is ready to be read from the radiation intensity storage circuit 64. The address information identifies the location from which to read the radiation intensity value for the second data interval. In another embodiment, the delta data processor 62 may accumulate the location information and communicate it along with the read signal either periodically or at the completion of a scanning operation. In still another embodiment, the radiation intensity values may be mapped into the radiation intensity storage circuit 64 in a manner such that the location information need not be communicated between the delta data processor 62 and the reconstruction processor 48. In this embodiment, the mapping of the radiation intensity storage circuit 64 is known to the reconstruction processor 48. Therefore, the reconstruction processor 48 only needs a read or ready signal from the delta data processor 62 or some other device indicating that either one or more radiation intensity values are stored or that the scanning operation is complete. When the next index pulse is detected by the data interval detector 80, the rotating gantry 14 has reached the fourth data interval and the address information is altered to identify storage locations for the “end data count” and “end time” for the third data interval and the “start data count” and “start time” for a fifth data interval. When the start of a first “pulse” on the pulse data signal is detected during the fourth data interval, the data counter 76 stores its current value in both the “end data count” location for the third data interval and the “start data count” location for the fifth data interval in the same manner as described above for the second/fourth data intervals during the third data interval. Likewise, the time counter 78 stores its current value in both the “end time” location for the third data interval and the “start time” location for the fifth data interval. At this point, the “start data count,” “end data count,” “start time,” and “end time” for the third data interval are now stored and the delta data controller 82 communicate a read signal and associated address information to the delta data processor 62 indicating such in the same manner as described above for the second data interval. The delta data processor 62 calculates a radiation intensity value for the third data interval and stores the radiation intensity value in the radiation intensity storage circuit 64 in the same manner as described above for the second data interval. When the next index pulse is detected by the data interval detector 80, the rotating gantry 14 has reached the fifth data interval and the address information is altered to identify storage locations for the “end data count” and “end time” for the fourth data interval and the “start data count” and “start time” for a sixth data interval. When the start of a first “pulse” on the pulse data signal is detected during the fifth data interval, the data counter 76 stores its current value in both the “end data count” location for the fourth data interval and the “start data count” location for the sixth data interval in the same manner as described above for the second/fourth data intervals during the third data interval. Likewise, the time counter 78 stores its current value in both the “end time” location for the fourth data interval and the “start time” location for the sixth data interval. At this point, the “start data count,” “end data count,” “start time,” and “end time” for the fourth data interval are now stored and the delta data controller 82 communicate a read signal and associated address information to the delta data processor 62 indicating such in the same manner as described above for the second data interval. The delta data processor 62 calculates a radiation intensity value for the fourth data interval and stores the radiation intensity value in the radiation intensity storage circuit 64 in the same manner as described above for the second data interval. The process described above for the second, third, and fourth data intervals is repeated for each data interval during scanning operations as the rotating gantry 14 turns. With reference to FIG. 3, the delta data storage circuit 60 includes a data storage block 84 and a time storage block 86. For the embodiment of the signal processor 46 described above, the data storage block 84 and the time storage block 86 each include four data storage locations. More specifically, the data storage block 84 includes data storage location A(0) 88, data storage location A(1) 90, data storage location A(2) 92, and data storage location B 94. The time storage block 86 includes time storage location A(0) 96, time storage location A(1) 98, time storage location A(2) 100, and time storage location B 102. With respect to the process described in reference to FIG. 2, the “start data count” for the second data interval is stored in data storage location A(0) 88 and the “start time” is stored in time storage location A(0) 96. The “end data count” for the second data interval is stored in data storage location B 94 and the “end time” in time storage location B 102. Similarly, the “start data count” for the third data interval is stored in data storage location A(1) 90 and the “start time” in time storage location A(1) 98. In the embodiment being described, the delta data processor 62 reads the “end data count” and “end time” for a measured data interval before the delta data controller 82 stores the “end data count” and “end time” for the next data interval. Therefore, the “end data count” for the third data interval is stored in data storage location B 94 and the “end time” in time storage location B 102. Likewise, the “start data count” for the fourth data interval is stored in data storage location A(2) 92, the “start time” in time storage location A(2) 100, the “end data count” in data storage location B 94, and the “end time” is in time storage location B 102. There are many ways of implementing the delta data storage circuit 60 and the associated method for storing and reading the time-based digital information representing the intensity of the detected radiation during a scanning operation. In one embodiment, the delta data storage circuit 60 is comprised of four sets of data (C) and time (T) storage locations (e.g., storage registers). The storage locations depicted in FIG. 3 are identified in these four sets as follows: CA(0) 88 and TA(0) 96, CA(1) 90 and TA(1) 98, CA(2) 92 and TA(2) 100, and CB 94 and TB 102. The contents of the data counter 76 and time counter 78 are transferred to one or more of the four pairs of storage locations as follows. On detection of a “pulse” on the pulse data signal by the data pulse detector 56, the current values in the count and time counters are transferred to: a) CA(0) and TA(0) for data intervals DI(1), DI(4), DI(7), etc., b) CA(1) and TA(1) for data intervals DI(2), DI(5), DI(8), etc., and c) CA(2) and TA(2) for data intervals DI(3), DI(6), DI(9), etc. On detection of the first “pulse” of each data interval, the current values of count and time counters are also transferred to CB and TB. This provides the time-based digital information necessary to determine the intensity of the detected radiation during the preceding data interval and the stored “start data count,” “end data count,” “start time,” and “end time” are read by the delta data processor 62. In the embodiment being described, the DATA and TIME measurements for data intervals DI(2), DI(3), DI(4), and DI(5) are calculated by the delta data processor 62 as follows: DATA(2) = CB − CA(0) TIME(2) = TB − TA(0) (DATA and TIME measurements for DI(2) are calculated before or at the end of DI(3).) DATA(3) = CB − CA(1) TIME(3) = TB − TA(1) (DATA and TIME measurements for DI(3) are calculated before or at the end of DI(4).) DATA(4) = CB − CA(2) TIME(4) = TB − TA(2) (DATA and TIME measurements for DI(4) are calculated before or at the end of DI(5).) DATA(5)=CB − CA(0) TIME(5) = TB − TA(0) (DATA and TIME measurements for DI(5) are calculated before or at the end of DI(6).) The following pseudo code performs the DATA and TIMEmeasurements described above: Initialize flagB = 0 for n=1:N while DI = n, upon receipt of a count pulse transfer counters to CA((n+2)(modulo3)) and TA((n+2)(modulo3)) if flagB == 0 transfer counters to CB and TB set flagB = 1 end end DATA(n−1) = CB−CA((n)(modulo3)) and TIME(n−1) = TB−TA((n)(modulo3)) reset flagB = 0 end With reference to FIG. 4, a sample timing diagram showing various signals within the embodiment of the CT scanner 10 described above during scanning operations. As shown, the timing diagram reflects signal levels during the first six data intervals (DI1-DI6) and the last data interval (DI N) of an exemplary detector. The INDEX signal represents the signal provided by the encoder 44 to the data interval detector 80 and defines the data intervals. As shown, the constant frequency of the pulses reflects the rotating gantry 14 moving at a constant velocity. The DATA signal (i.e., offset data signal) represents the signal provided by the summing module 66 to the IFC 70 produced by combining the analog data signal from the radiation detector 40, 42 with the offset current from the offset module 68. The IFC signal (i.e., pulse data signal) represents the pulse train output of the IFC 70 that is provided to the data pulse detector 72 and data counter 76. The OSC signal represents the free-running output of the oscillator 74 provided to the time counter 78. The resolution of the diagram does not permit identification of the frequency of the OSC signal. Nevertheless, the frequency of the pulses in the OSC signal is relatively constant at a predetermined very high frequency. The STORE A signal represents the signal from the delta data controller 82 directing the data counter 76 and time counter 78 to store current values in start locations within the delta data storage circuit 60. Note that the STORE A signal is communicated each time a pulse is detected on the IFC signal during each data interval. The STORE A signal operates in conjunction with address information provided by the delta data controller 82 to the delta data storage circuit 60 to store the values of the counters in selected “start data count” and “start time” storage locations for measurement of the succeeding data interval. The STORE B signal represents the signal from the delta data controller 82 directing the data counter 76 and time counter 78 to store current values in end locations within the delta data storage circuit 60. Note that the STORE B signal is communicated on the first pulse detected on the IFC signal during each data interval. The STORE B signal operates in conjunction with address information provided by the delta data controller 82 to the delta data storage circuit 60 to store the values of the counters in selected “end data count” and “end time” storage locations for measurement of the preceding data interval. A pair of STORE A and STORE B signals identify the boundaries of the measurement period for a measured data interval. Note that the measurement period for a measured data interval starts in the preceding data interval when the last “pulse” on the DATA signal is detected and ends in the succeeding data interval when the first “pulse” on the DATA signal is detected. This can be seen by comparing the measurement periods to the time for the data interval, e.g., MP 2 to DI 2, MP 3 to DI 3, etc. The READ signal represents the signal from the delta data controller 82 to the delta data processor 62 indicating that the “start data count,” “end data count,” “start time,” and “end time” are stored for a measured data interval. The delta data processor 62 uses address information provided by the delta data controller 82 in conjunction with the READ signal to read the time-based digital information stored for the data interval. In another embodiment of the method for storing and reading the time-based digital information, the offset current provided by the offset module 68 is reduced to a point where the delta data channel circuit 56a guarantees a “pulse” on the pulse data signal at least once during every 2½ data intervals. In this embodiment, the delta data storage circuit 60 is comprised of six pairs of data (C) and time (T) storage locations (e.g., storage registers). Four of these pairs are used to store “start data count” and “start time count” and two are used to store “end data count” and “end time count.” The measurement periods can extend up to two data intervals preceding and two data intervals succeeding the measured data interval. That is, the start pulse may be two intervals before the measured data interval and the stop pulse may be two intervals after the measured data interval. The measurement period for a measured data interval starts with the last pulse preceding the measured data interval (i.e., within the two preceding data intervals) and terminates with the first pulse following the measured data interval (i.e., within the two succeeding data intervals). If no pulse is produced within the two preceding data intervals, the measurement period starts with the first pulse of the measured data interval or, if no pulse is produced within the two succeeding data intervals, the measurement period ends with the last pulse of the measured data interval. Data intervals can overlap by a greater amount at high attenuation, although on the average, the measurements will be centered on the current or measured data interval. With reference to FIG. 5, a timing diagram shows various scenarios (i.e., scenarios a through i) for measurements periods associated with a measured data interval (n) for the embodiment having a “pulse” on the pulse data signal as least once during every 2½ data intervals. Two preceding data intervals are identified as n−2 and n−1. Two succeeding data intervals are identified as n+1 and n+2. If one or more data pulses are detected and the data interval includes a start or end boundary for a measurement period, a first line 104 is identified in the scenario. If two or more data pulses are detected and the data interval includes the start and end boundary for a measurement period, a second line 106 is identified in the scenario. If data pulses are not detected during a data interval, a blank data interval 108 is identified in the scenario. If it does not matter whether or not data pulses are detected during a data interval, a dashed line 110 is identified in the scenario. A measurement period with start and stop boundaries is identified by bracket 112 in each scenario. If no data pulses are detected for three consecutive data intervals, an error condition exists for this embodiment. Scenario i creates a situation in which two or more data pulses are required during data interval n. Otherwise, only one pulse is required in a data interval used as a start or end boundary for the measurement period. In an embodiment using storage locations, the storage location are identified in these six sets as follows CA(0) and TA(0), CA(1) and TA(1), CA(2) and TA(2), CA(3) and TA(3), CB(0) and TB(0), and CB(1) and TB(1). In general, the contents of the data counter 76 and time counter 78 are transferred to one or more of the four pairs of storage locations as follows. On detection of a “pulse” on the pulse data signal by the data pulse detector 72, the contents of the data counter 76 and time counter 78 may, for example, be transferred to: a) CA(0) and TA(0) for data intervals DI(3), DI(7), DI(11), etc., b) CA(1) and TA(1) for data intervals DI(4), DI(8), DI(12), etc., c) CA(2) and TA(2) for data intervals DI(5), DI(9), DI(13), etc., and d) CA(3) and TA(3) for data intervals DI(6), DI(10), DI(14), etc.The pseudo code below for the embodiment with six pairs of data (C) and time (T) storage locations identifies additional storage combinations. On detection of the first pulse within a data interval, the contents of the data counter 76 and time counter 78 may, for example, be transferred to: a) CB(0) and TB(0) for data intervals DI(3), DI(5), DI(7), etc. and b) CB(1) and TB(1) for data intervals DI(4), DI(6), DI(8), etc.The pseudo code below for the embodiment with six pairs of data (C) and time (T) storage locations identifies additional storage combinations. This provides the time-based digital information necessary to determine the intensity of the detected radiation during a data interval for the embodiment being described. The stored “start data count,” “end data count,” “start time,” and “end time” are read by the delta data processor 62 at the end of the second succeeding data interval. The following pseudo code performs the DATA and TIME measurements for the embodiment described above with reference to FIG. 5: initialize flagB(0) = 0; flagB(1) = 1for n = 1:N while DI = n, upon receipt of a count pulse transfer counters to CA((n+1)(modulo4)) and TA((n+1)(modulo4)) if flagB(0) == 0 transfer counters to CB(0) and TB(0) end if flagB(1) == 0 transfer counters to CB(1) and TB(1) end if flagB(0) AND flagB(1) == 0 transfer counters to CA((n)(modulo4)) and TA((n)(modulo4)) end set flagB(0)=1 set flagB(1)=1 end COUNT(n−2) = CB(n(modulo2)) − CA(n+2(modulo4)) TIME(n−2) = TB(n(modulo2)) − TA(n+2(modulo4)) transfer counters to CA((n+2)(modulo4)) and TA(n−2(modulo4)) and to CB(n(modulo2)) and TB (n(modulo2)) reset flagB(n(modulo2)) = 0end In summary, the various embodiments described above provide what may be referred to as a symmetrical delta data mode for measuring the intensity of detected radiation for data intervals during scanning operations in a CT scanner. The symmetrical delta data mode produces a measurement period for a measured data interval that extends into both the preceding and succeeding data intervals. On the average, the measurement period is centered on the measured data interval, thus producing an average skew of zero. As a result, artifacts due to data skewing are reduced from those of previous delta data modes. Moreover, under conditions of high attenuation, the measurement period is significantly longer than the data interval thus producing a more integrated signal and reducing quantum noise, thereby thus increasing the dynamic range of the overall. The increase in measurement period as the input signal decreases produces an adaptive filtering effect in the analog domain that can potentially improve image quality more effectively than subsequent filtering in the digital domain. In various alternate embodiments, the offset current can be reduced to less than one “pulse” in the pulse data signal per data interval. The reduction in the offset current decreases shot noise associated with the offset current. In addition, reducing the offset current decreases the effects of quantization noise and 1/f noise. The resulting overall noise reduction improves image quality and can significantly extend the system dynamic range. While the invention is described herein in conjunction with exemplary embodiments, it is evident that many alternatives, modifications, and variations will be apparent to those skilled in the art. Accordingly, the embodiments of the invention in the preceding description are intended to be illustrative, rather than limiting, of the spirit and scope of the invention. More specifically, it is intended that the invention embrace all alternatives, modifications, and variations of the exemplary embodiments described herein that fall within the spirit and scope of the appended claims or the equivalents thereof.
	claims	1. A fuel assembly for being loaded into a reactor core wherein a control rod-side water gap width and an opposite-side water gap width are almost equal to each other, comprising: a fuel bundle having a plurality of fuel rods arranged in a square lattice pattern and at least one neutron moderator rod, each fuel rod being filled with nuclear fuel pellets; an upper tie plate and a lower tie plate holding upper end portions and lower end portions of said fuel rods respectively; and means for fixing a channel box covering said fuel bundle to one corner of said upper tie plate by a channel fastener, a center of said at least one neutron moderator rod being shifted toward said one corner, away from a cross sectional center of the fuel assembly, wherein said plurality of fuel rods include a plurality of short-length fuel rods having a shorter active fuel length than remaining fuel rods, the number of said short-length fuel rods being arranged in one diagonally divided half area opposite to said one corner is larger than that in the other diagonally divided half area. 2. A fuel assembly according to claim 1 , wherein the fuel assembly is divided into a first region including said one corner and a second region by a diagonal line in a cross section, and an average uranium enrichment of the fuel rods in said second region is higher than that of the fuel rods in said first region. claim 1
041479386	summary	FIELD OF THE INVENTION The invention relates to a bimetallic band device which improves the fire resistance of a nuclear fuel cask. BACKGROUND Spent nuclear fuel elements, being radioactive and generating significant thermal energy, require special containers or casks for storage and transportation from a reactor to a reprocessing or storage site. A number of such spent fuel shipping casks have been designed. Typical construction, as exemplified by U.S. Pat. No. 3,113,215 to Allen, includes a central cavity, an inner container or shell, a radiation shielding filler, an outer container or shell, and heat rejecting fins projecting outwardly from the outer shell. Federal regulations currently require that a spent fuel shipping cask survive an 802.degree. C. fire for 1/2 hour. The heat rejecting fins, which normally serve to conduct heat away from the cask interior, may then conduct heat inwardly. This, of course, is undesirable as it reduces the ability of a cask to withstand a fire. One proposed method of fire protection is that of extinguishing the fire by liquids, foams, or gases. If such extinguishing agents are contained within the cask itself, it may be difficult to provide an effective amount of the agent. On the other hand, if the extinguishing agent is located within the carrier used to transport the cask, the extinguishing agent may become ineffective if the cask and carrier become separated. A method disclosed by U.S. Pat. No. 3,414,727 to Bonilla would cool the cask surrounding the radiation shield with a safety shield of material which melts at a temperature lower than the radiation sheild and adapted to flow out of its enclosed space when subjected to external heat. Another concept is to incorporate a hydrated substance within the cask. Exposure to an exterior heat source would cause dissociation and vaporization of the contained water; heat would be rejected as latent and sensible heat as the resultant water vapor was vented from the cask. Examples of such substances are: hydrated calcium sulfate (plaster) as disclosed in U.S. Pat. No. 3,466,662 to Blum or hydrated aluminum and iron oxides as disclosed by U.S. Pat. No. 3,780,309 to Bochard. These methods suffer the disadvantage that a limit to the amount of heat that may be rejected is set by the amount of hydrated material that is contained within the cask. Furthermore, if the water within the hydrated substance is relied upon for neutron shielding, that shielding is degraded upon exposure to fire. A related concept, as disclosed by U.S. Pat. No. 3,737,060 to Blum would place neutron shielding such as borated wood or aluminous cement on the exterior of the outer shell and between the heat rejecting fins. This arrangement presumably would afford some fire protection through charring of the wood or dehydration of the cememt; however it would again suffer the disadvantages noted above. It is known, especially in the art of designing skins for spacecraft, that certain bimetallic devices may be used for controlling the spacecraft skin temperature by controlling the emissivity of that surface (see, for example, U.S. Pat. Nos. 3,205,937, 3,220,647, 3,307,783, 3,362,467, and 3,411,156). In general, those devices are designed to operate in the cold vacuum of outer space and are effective for radiative heat transfer. The present invention is designed to operate at atmospheric pressure and temperatures ranging from ambient to those found in flames. The present invention is effective in controlling heat transfer by conduction and convection as well as by radiation. Ablative or intumescent materials could be applied to the surface of the cask, but these materials could interfere with normal fin heat rejection. Thermal isolation of the cask may be accomplished by irreversibly decoupling the fins from the cask; for example, by melting the fin substructure or by the formation of cavities near the fin surface. These schemes, along with most of the other proposed methods, suffer the disadvantage of irreversibility. Generally, after a fire wanes, human intervention would be required to restore the thermal conductivity of the cask and thereby prevent excessive temperature rise from internal heat generation. Depending on accident conditions, such human intervention may not be possible. SUMMARY OF THE INVENTION In accordance with the present invention there is provided a plurality of metallic bands, individually located between adjacent cooling fins of a nuclear fuel cask. Each band may be comprised of strips of two dissimilar metals bonded together along a side and formed to the general contour of the cask each band encircling or girdling the cask. Upon heating such as in an accidental fire, the bands expand to block normal heat transfer between the fins and the environment. One object of the present invention is to provide improved fire protecton for nuclear fuel casks. Another object of the present invention is to provide fire protection for nuclear fuel casks which is fully automatic and does not require human intervention for deployment. One advantage of the present invention is that it is fully reversible, with the cask automatically returning to its prefire configuration following extinguishing of the fire. Another advantage of the present invention is that quantities of consumable fire extinguishing agents or heat removal agents are not required for its operation.
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	abstract	A device for, and method of manufacture of, a focused anti-scatter grid for improving the image contrast of x-ray images produced in medical, veterinary or industrial applications. The grid comprising a series of modular units so juxtaposed with each other as to form a series of focused channels for the passage of the focused imaging x-rays. The modules comprise a series of focusing ribbons of a heavy metal or a series of mating solid arcuate forms, formed of a polymer and having on at least one side surface a layer of heavy metal.
	description	This is a continuation, under 35 U.S.C. §120, of copending International Application PCT/EP2014/058721, filed Apr. 29, 2014, which designated the United States; this application also claims the priority, under 35 U.S.C. §119, of German Application DE 10 2013 214 230.7, filed Jul. 19, 2013; the prior applications are herewith incorporated by reference in their entirety. In a nuclear power station, in the event of situations involving incidents or accidents, a possibly significant release of radioactive fission products, in particular iodine, aerosols and noble gases, must be anticipated in accordance with the incident and any counter-measures initiated. As a result of leakages of the containment, in this instance there must also be assumed a release and distribution of activity in the power station buildings (for example, auxiliary plant building, switching installation, control room, etc.) before a release into the power station environment is brought about. In this instance, in particular the release of noble gases is a problem for the power station operators in addition to the release of activity connected with aerosols. Under some circumstances, there is also produced a massive release of noble gases during the introduction of a filtered pressure release and the formation of a noble gas cloud above the power station site. Depending on the weather conditions, longer-term pollution cannot be completely excluded. In order to introduce so-called accident management measures, it is absolutely necessary for the conditions in the control area, which is also referred to as a control room or management location, to allow the operators to remain without an inadmissible radiation exposure and contamination of the operators being produced. In the event of configuration-exceeding incidents with “Station Black-Out” (SBO), the ventilation systems and filter systems operating normally and in accordance with provisions are no longer available in order to ensure the significant technical ventilation parameters to maintain the accessibility of the control room. Previous concepts make provision for the control room to be isolated in order to control such scenarios. The supply is brought about, for example, with mobile ventilation systems which are provided with different filters. A satisfactory retention of noble gases is not possible with those systems. Other concepts supply the control room with stored compressed air. However, the storage in pressurized containers for a longer time is very complex and is therefore limited. A modular and mobile system construction is not practically possible. Furthermore, pressure storage concepts require a high level of complexity in the case of retrofitting in operational plants. It is accordingly an object of the invention to provide a ventilation system and an associated operating method for use during a serious incident in a nuclear plant, which overcome the hereinafore-mentioned disadvantages of the heretofore-known systems and methods of this general type and in which the ventilation system is kept as small and compact as possible for a control room or operation center of a nuclear plant or a similar room which is accessible to operators and in which the ventilation system allows a supply of decontaminated fresh air at least for a time of a few hours in the event of serious incidents with the release of radioactive activity so that there is produced the smallest possible radioactive exposure of operators who are present in the control room. In this instance, the proportion of radioactive noble gases in the fresh air supplied to the control room is particularly intended to be as small as possible. The ventilation system is further intended to have as passive a character as possible and to consume only a small amount of electrical energy. With the foregoing and other objects in view there is provided, in accordance with the invention, a ventilation system for an operator-accessible operations room in a nuclear plant, in particular a control room in a nuclear power station. The ventilation system comprises an external inlet, an air supply line guided from the external inlet to the operations room, a first fan connected to the air supply line and a first noble gas adsorber column connected to the air supply line. There is also provided an external outlet, an air discharge line guided from the operations room to the external outlet, a second fan connected to the air discharge line and a second noble gas adsorber column connected to the air discharge line. A switch-over device exchanges functions of the first and second noble gas adsorber columns. With the objects of the invention in view, there is also provided a method for operating a ventilation system for an operator-accessible operations room in a nuclear plant, in particular a control room in a nuclear power station. The method comprises guiding supply air through one of the noble gas adsorber columns thereby charging the one noble gas adsorber column with radioactive noble gases while simultaneously guiding discharge air through the other of the noble gas adsorber columns and thereby backwashing the other noble gas adsorber column. Advantageous embodiments are set out below and will be further appreciated from the following detailed description. The ventilation system according to the invention advantageously has inter alia an aerosol and iodine filtering module. In this instance, the intake air in the air supply line is drawn in through a fan and guided through high-efficiency particulate air filters in order to separate the aerosols. After the separation of the airborne particulates, radioactive iodine compounds are advantageously separated in an activated carbon filter bed. Impregnated activated carbon can be used in order to separate the radioactive methyl iodide by using isotope exchange or salt formation. A particulate filter is advantageously connected downstream of the activated carbon bed in order to retain dust particles. The air which is filtered in this manner is then supplied to a noble gas module in a second process step. The noble gas module substantially contains two adsorber columns in a twin configuration which are filled with adsorbent(s), preferably activated carbon. The adsorbent of the columns may also be constructed from a plurality of layers of activated carbon and/or zeolite and/or molecular sieves. The air supply is introduced into the first adsorber column, wherein the noble gases such as, for example, xenon, krypton, are decelerated by a dynamic adsorption during the passage thereof through the column. After the column, a filter is advantageously disposed to retain adsorber particulates. The discharge air from the room region to be supplied is simultaneously guided through the second adsorber column and brings about at that location a backwashing of the previously accumulated noble gas activity so that this column is again ready for charging after the change-over. The change-over is carried out at the latest shortly before the cessation of the activity in the first adsorber column, wherein it is then backwashed with the discharge air. The change-over is preferably triggered passively by a timing member or an activity measurement unit. The backwashing is advantageously supported by a fan in the air discharge line, wherein the volume increase of the discharge air flow as a result of the reduced pressure increases the backwashing process of the noble gases. There is advantageously located in the air discharge line of the control room a throttle which results in the passive overheating of the discharge air and therefore a reduction of the moisture which is located in the discharge air (expansion drying). The desorption speed of the noble gases in the downstream adsorber column to be flushed is thereby promoted. A throttle and/or an air dryer are advantageously located in the air supply line to the noble gas module in order to prevent excessively high moisture being conveyed to the noble gas columns. The noble gas module can further be provided with a passive cold storage device in order to increase the k values. The k value describes in this context the adsorption capacity of the adsorber material for noble gas in, for example, the unit cm3 of noble gas/g of adsorbent. The k value is dependent on the temperature, pressure and moisture content of the gas. It is generally established empirically. The adsorber columns are preferably operated with the pressure changing method, that is to say, reduced pressure of the column to be flushed and excess pressure of the column to be charged (in relation to atmospheric pressure in each case) in order to improve the k values of the columns and to reduce the dimensions thereof. The excess pressure in the adsorber column through which the air supply flows is, for example, regulated with an adjustment valve in the air supply line. The discharge air is discharged together with the backwashed noble gases into the power station environment with sufficient spacing from the air supply intake. The ventilation system advantageously includes a control unit and corresponding adjustment members for through flow and pressures. The advantages obtained with the invention particularly involve the radioactive noble gases being simultaneously retained from the supply air of the control room in addition to the air-borne activities in the form of aerosols and iodine/iodine compounds (in particular organo-iodine). Even long-life noble gas isotopes such as krypton 85 can be reliably separated from the air supply flow with the pressure change and flushing method of the twin columns. The conditions necessary for removing the noble gases from the sorbent/adsorbent are supported passively by expansion overheating. There is a requirement for electrical operating current substantially only for the fans in the air supply and the air discharge line and to a small extent for the associated control unit and for the switching device for switching between the operating cycles. That requirement can readily be met for at least 72 hours with an autonomous energy supply module (for example, by batteries and/or a diesel generator unit). In summary, the following functions are provided in order to ensure the accessibility of the control room: isolation of the control room air from the remaining parts of the building, excess pressure with respect to the adjacent building rooms (for example, <1 mbar), compliance with the admissible carbon monoxide and carbon dioxide concentration, iodine retention, aerosol retention, retention of noble gases (for example, Kr, Xe), limiting of the dose (for example, <100 mSv/7d), temperature limiting in order to comply with the I&C temperature qualifications, ensuring the above-mentioned functions for at least 72 hours. Other advantages are summarized as key points: modular and mobile system construction, low complexity and high flexibility for integration in current systems, low maintenance complexity, a complex storage of breathable air is unnecessary, it is possible to cover relatively large quantities of air (air exchange) and spatial regions. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a ventilation system and an associated operating method for use during a serious incident in a nuclear power plant, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. Referring now in detail to the single FIGURE of the drawing, there is seen an incident ventilation system which is referred to briefly as a ventilation system 2 that is used for supplying fresh air for a control room or operation center 4 (also referred to as a Main Control Room (MCR)) of a nuclear power station 6 in situations involving accidents or incidents, in particular during the start phase of a serious incident with a release of nuclear fission products within the power station building and where applicable also to the environment. In such scenarios, which are generally associated with the failure of the individual power supply of the nuclear power station 6 and therefore also with the failure of the normally operational ventilation system (not illustrated) for the control room 4, it is particularly important to still be able to keep the control room 4 occupied for a specific time—for instance, up to 72 hours after the start of the incident—without endangering the operators in order to initiate initial counter-measures and to monitor them. Possibly, the operators may also have to remain in the control room 4 until secure evacuation is possible after an initial activity maximum has cooled. For this purpose, the ventilation system 2 for the control room 4 is configured, on one hand, for a supply of decontaminated and oxygen-rich fresh air—also referred to as supply air—from the environment of the control room 4 or the power station building and provided with corresponding filter and cleaning stages. On the other hand, the ventilation system 2 brings about a discharge of consumed air rich in carbon dioxide—also referred to as discharge air—from the control room 4 into the environment. In this instance, unlike other, previously conventional concepts, neither a fresh air supply from an associated compressed air storage system nor a substantial recirculation and re-processing of the air in the inner space of the control room 4 is provided. In specific terms, an air supply line 10, which is also referred to as a fresh air supply line or, in brief, as a fresh air line and through which fresh air from the environment is drawn in by using a fan 12 during the operation of the ventilation system 2 and is conveyed into an inner space 8 of the control room 4, is connected to the inner space 8 which is at least approximately hermetically encapsulated with respect to the outer environment. An intake inlet or, in brief, an inlet 14 of the air supply line 10 can be located at a spacing from the control room 4, in particular outside the power station building. Depending on the progress of the incident, the fresh air drawn in through the inlet 14 can nevertheless be substantially charged with radioactive fission products, in particular in the form of aerosols, iodine and iodine compounds and noble gases. Those components are intended to be removed as completely and reliably as possible from the fresh air flow—also referred to as an air supply flow—before the flow is introduced through a conduit 16 in an enclosure wall 18 (only illustrated as a cutout) into the inner space 8 of the control room 4. To this end, downstream of the inlet 14 when viewed in the direction of the fresh air flow, a first filter stage in the form of an aerosol filter 20 is connected to the air supply line 10. In this instance, by way of example, the aerosol filter 20 is produced by two HEPA filters 22 which are connected in parallel in terms of flow (HEPA=High Efficiency Particulate Air filter). The HEPA filters 22 accordingly bring about a highly efficient separation of the aerosol particulates (also referred to as airborne particulates) from the fresh air flow, in particular in relation to the isotopes Te, Cs, Ba, Ru, Ce, La. Further downstream, a second filter stage having an iodine filter 24 and a downstream particulate filter 26 is connected to the air supply line 10. The iodine filter 24 is preferably in the form of an activated carbon filter bed having a layer thickness of, for example, from 0.1 to 0.5 m. After the separation of the airborne particulates as carried out previously in the aerosol filter 20, radioactive iodine compounds and elemental iodine are separated in the iodine filter 24, for example, at a k value >8 for contact times of from 0.1 to 0.5 s. In order to separate the radioactive methyl iodide by using isotope exchange or salt formation, impregnated activated carbon (for example, with potassium iodide as the impregnation device) can be used. The particulate filter 26, which is connected downstream of the iodine filter 24, is provided for retaining dust particles from the activated carbon bed. Downstream of the second filter stage, a conveyor fan or in brief the fan 12 is connected to transport the fresh air flow into the air supply line 10. The preferably electrically driven fan 12 has a suction power in the range, for example, of from 1000 to 6000 m3/h. In order to provide the necessary operating current, there is provided an autonomous power supply module 28 which is independent of the normally operational individual power supply and preferably also of the conventional emergency power network (across the plant), for example, on the basis of electrical batteries/accumulators and/or a diesel generator unit. The power supply module 28 becomes activated as required, preferably independently in the manner of a non-interrupted power supply, or is alternatively controlled through an associated control unit 30. Further downstream there may optionally be connected to the air supply line 10 an air dryer 32 which is also referred to as a cold trap and with which condensable components can be separated from the fresh air flow. This may be, for example, a passive cold trap with silica gel and/or ice as a drying agent. The moisture content of the fresh air flow which flows into the downstream functional units (see below) is thereby reduced. An alternatively or additionally present throttle 34, which is disposed in this case in the embodiment when viewed in the direction of flow of the fresh air downstream of the air dryer 32 and which acts on the fresh air flow in accordance with the principle of expansion drying, serves the same purpose. The throttle may be, in particular, an adjustable throttle valve. Following the filtering and drying, the fresh air flow flows, for a corresponding position of associated positioning members (see below), for example, through a line portion 36, to which a noble gas adsorber column or, in brief, an adsorber column 38 is connected. In this instance, the noble gases which are contained in the fresh air flow, in particular xenon and krypton, are bound, in the context of a dynamically adjusting equilibrium by physical and/or chemical adsorption, to the adsorbent present in the adsorber column 38, and consequently decelerated in the line portion 36 as long as the adsorption capacity of the adsorber column 38 is not yet exhausted. In particular one or more layers of activated carbon and/or zeolite and/or molecular sieves may be provided as the adsorbent. A line portion which leads to the control room 4 and to which a particulate filter 40 is connected in order to retain loosened adsorber particulates, is connected downstream of the adsorber column 38. Finally, the fresh air flow which is decontaminated in the manner described is introduced through the conduit 16 through the enclosure wall 18 of the control room 4 into the inner space 8 thereof so that non-consumed, oxygen-rich air for breathing with an activity degree which is permitted for the operators is supplied thereto. The air exchange is brought about by the discharge of consumed, carbon-dioxide-rich air for breathing from the control room 4 through an air discharge line 44, which is connected to the inner space 8 thereof and which is directed through a conduit 42 in the enclosure wall 18 into the environment and to which a fan 46 is connected in order to support the gas transport. The fan is preferably an electrically driven fan 46 which is supplied with electric current similarly to the fan 12 by the power supply module 28. Since the adsorption capacity of the adsorber column 38 which acts on the fresh air flow is generally already exhausted after a relatively short operating time for a practicable construction size, the ventilation system 2 is configured for a backwashing of the adsorbed noble gases into the environment during current operation. For this purpose, there are provided two substantially structurally identical adsorber columns 38 and 48 which are acted on through corresponding line branches and connections and positioning members, in this instance in the form of 3-way valves, with fresh air or discharge air in such a manner that one of the two adsorber columns 38 and 48, as already described, acts on the fresh air flow during adsorption operation, while the other is simultaneously backwashed during desorption operation or flushing operation by the discharge air flow and is consequently made ready for the next adsorption cycle. The function of the adsorber columns 38 and 48 can be transposed and consequently a change can be brought about in relation to the respective column cyclically between adsorption operation and desorption operation by switching over the positioning members. In the embodiment illustrated in the FIGURE, this functionality is brought about in that one adsorber column 38 is disposed in the line portion 36 and the other adsorber column 48 is disposed in a line portion 50 with a non-parallel connection in flow terms. The two line portions 36 and 50 are combined at one side in a 3-way valve 52 and at the other side in a union 54 which is disposed at the intake side of the fan 46. Furthermore, at one side between the 3-way valve 52 and the two adsorber columns 38, 48, a transverse connection 60 which can be switched by two 3-way valves 56 and 58 is connected between the two line portions 36 and 50 and is connected through a T-connection 62 to the portion of the air supply line 10 leading to the particulate filter 40. At the other side, in a similar configuration, a transverse connection 68 which can be switched by two 3-way valves 64 and 66 is connected between the adsorber columns 38, 48 and the union 54 and is connected through a T-connection 70 to the portion of the air supply line 10 coming from the throttle 34. In the case of correspondingly selected valve positions, as already described above, the supply air from the throttle 34 flows through the T-connection 70, the 3-way valve 66, the adsorber column 38 at the bottom in the FIGURE, the 3-way valve 58 and the T-connection 62 to the particulate filter 40 and, from there, further to the control room 4. In the other line strand, the discharge air from the control room 4 flows through the 3-way valve 52, the 3-way valve 56, the adsorber column 48 at the top in the FIGURE and the 3-way valve 64 to the suction connection of the fan 46 and, from there, further to a discharge chimney or to another outlet 72, which is advantageously located with spacing from the inlet 14 for fresh air. That is to say, the noble gases which are accumulated in the previous cycle in the adsorber column 48 by adsorption are desorbed in this operating mode by the substantially noble-gas-free discharge air from the inner space 8 of the control room 4 by the adsorbent and backwashed with the discharge air flow into the environment. The backwashing is supported by the fan 46 which is disposed downstream of the backwashed adsorber column 48, wherein the volume increase of the discharge air flow as a result of the reduced pressure increases the backwashing process of the noble gases. There is located, in the air discharge line 44 of the control room, when viewed in the direction of the discharge air flow, upstream of the 3-way valve 52 and consequently upstream of the adsorber column 48 presently being used for flushing operation, a throttle 74, preferably in the form of an adjustable throttle valve which results in the passive overheating of the discharge air and therefore a reduction of the moisture located in the discharge air (expansion drying). The desorption speed of the noble gases in the downstream adsorber column 48 is thereby promoted. After the switch-over, the functions of the adsorber columns 38 and 48 are transposed. Now the fresh air flows from the throttle 34 through the 3-way valve 64, the adsorber column 48 and the 3-way valve 56 to the particulate filter 40 and, from there, to the control room 4. However, the discharge air from the control room 4 flows from the throttle 74 through the 3-way valve 52, the 3-way valve 58, the adsorber column 38 and the 3-way valve 66 to the fan 46 and, from there, to the outlet 72. The previously charged adsorber column 38 is now backwashed by the discharge air while the adsorber column 48 is available for cleaning the fresh air and accordingly for repeated charging. In order to control the switch-over operations by using the 3-way valves 52, 56, 58, 64, 66, there is provided the control unit 30 which advantageously also controls the two fans 12 and 46 and optionally other positioning members for throughflow and pressures. It will be self-evident to the person skilled in the art that the switch-over functionality can also be brought about by using other line topologies and positioning members in an equivalent manner. As indicated by the broken peripheral lines, the ventilation system 2 is preferably constructed in a modular manner from a noble gas module 76, an iodine and aerosol module 78 and a power supply module 28. The boundaries between the modules can naturally in detail also be selected to be different and there may be other modules or sub-modules. The individual modules are received, for example, in standard containers in a transportable manner so that simple transport to the installation location and simple construction at that location can be carried out by connecting the associated, standardized line connections. Even if the description was previously directed towards the ventilation of the (central) control room of a nuclear power station, it is nevertheless clear that the ventilation system 2 can also be used for incident ventilation of other areas within a nuclear power station or more generally a nuclear plant—for instance, also combustion element storage areas, reprocessing plants, fuel processing plants, etc.—for example, of auxiliary plant buildings, switching plant rooms, measurement control rooms or other operating and monitoring rooms. The designation “operations room” is also used in a summarizing manner as a keyword for such rooms.
	claims	1. A method of forming a protection layer on a wafer slice to form a specimen for a transmission electron microscope (TEM) inspection, the method comprising:coating the wafer slice with a protection material on top, bottom and side surfaces of the wafer slice, wherein one of the side surfaces comprises a beveled side surface comprising an inspection point; and,compressing the protection material to the wafer slice. 2. The method of claim 1, further comprising:heating a stage adapted to support the wafer slice prior to coating the wafer slice with the protection material; and,positioning the wafer slice on an upper surface of the stage. 3. The method of claim 2, further comprising:coating the upper surface of stage with the protection material before positioning the wafer slice. 4. The method of claim 3, wherein compressing the protection material comprises:providing a covering member adapted to fit over the top surface of the wafer slice and around the side surfaces of the wafer slice following the step of coating the wafer slice with the protection material. 5. The method of claim 4, wherein compressing the protection material further comprises:compressing the protection material to the wafer slice using the covering member. 6. The method of claim 5, further comprising:cooling the stage after compressing the protection material. 7. The method of claim 1, wherein the protection material comprises an epoxy resin compound. 8. The method of claim 4, wherein the covering member is formed in accordance with a shape of the wafer slice. 9. A method of forming a specimen for a transmission electron microscope (TEM) inspection, the method comprising:cutting a wafer slice from a wafer, the wafer slice comprising a beveled side surface comprising an inspection point;forming a protection layer on the wafer slice protecting the inspection point;forming a first preliminary specimen by cutting the wafer slice, wherein the first preliminary specimen comprises the inspection point;forming a second preliminary specimen by grinding top and bottom surfaces of the first preliminary specimen, wherein the second preliminary specimen comprises the inspection point; and,forming a specimen by etching portions of top and bottom surfaces of the second preliminary specimen around the inspection point to respective predetermined depths. 10. The method of claim 9 wherein forming the protective layer on the wafer slice comprises:coating the wafer slice with a protection material on top, bottom and side surfaces, wherein one of the side surfaces comprises the beveled side surface; and,compressing the protection material to the wafer slice. 11. The method of claim 10, further comprising:heating a stage adapted to support the wafer slice prior to coating with the protection material; and,positioning the wafer slice on an upper surface of the stage. 12. The method of claim 11, further comprising:coating the upper surface of stage with the protection material before positioning the wafer slice. 13. The method of claim 12, wherein compressing the protection material comprises:providing a covering member adapted to fit over the top surface of the wafer slice and around the side surfaces of the wafer slice following the step of coating the wafer slice with the protection material. 14. The method of claim 13, wherein compressing the protection material further comprises:compressing the protection material to the wafer slice using the covering member. 15. The method of claim 14, further comprising:cooling the stage after compressing the protection material. 16. The method of claim 9, wherein the protection material comprises an epoxy resin compound. 17. The method of claim 13, wherein the covering member is formed in accordance with a shape of the wafer slice. 18. The method of claim 9, wherein forming the second preliminary specimen comprises:securing the first preliminary specimen to a stage such that the inspection point faces in a direction parallel to the surface of the stage; and,grinding the first preliminary specimen. 19. The method of claim 9, wherein the second preliminary specimen is ground while the inspection point is viewed.
056125437	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS Reference will now be made in detail to the present preferred embodiments of the invention as illustrated in the accompanying drawings. In accordance with the present invention, there is provided a basket for transporting, storing, and containing nuclear fuel assemblies having an assembly of sleeves with a plurality of sleeves arranged in a uniform pattern and secured within a cylindrical shell. Each of the plurality of independent sleeves being sized to secure and contain a fuel assembly. A plurality of alternating sleeves of the plurality of independent sleeves are configured to include an angular shaped separator element secured to each corner of each of the plurality of alternating sleeves. A sheet of neutron absorbing material is positioned between each of the plurality of alternating sleeves for maintaining fission reactions within the basket below a critical level necessary to sustain a fission reaction. A support element for positioning and securing the plurality of independent sleeves is secured within the cylindrical shell. A bottom plate is secured to the bottom of the cylindrical shell providing vertical support for the plurality of independent sleeves. A shield lid is secured to the cylindrical shell and includes a plurality of disc elements and an access port for selective entry into the basket and a lid element is secured to the shield lid and to the cylindrical shell. The lid element including an access port for selective entry into the basket. In FIG. 1, the multi-purpose sealed boiling water reactor fuel basket 10 is shown with shell 12 having a top end 14, a bottom end 16, an outer wall 18 and an inner wall 20, according to a preferred embodiment of the invention. Shell 12 is preferably cylindrically configured but may be provided in other geometric configurations if desired, such as circular, square, rectangular, or the like. Basket 10 is preferably composed of a durable, resilient, non-corrosive material such as steel or steel alloys, and is typically shipped or transported in a transportation, storage, or shipping cask commonly used in the art. As seen in FIG. 1, basket 10 includes an assembly of independent sleeves 22 comprising a plurality of independent sleeves 24. Sleeves are preferably configured having a square cross section and positioned and secured in a uniform pattern inside shell 12 which is preferably cylindrically shaped. Each sleeve 24 is preferably sized and shaped to contain one boiling water reactor fuel assembly 48, however, in alternative embodiments fuel assemblies for different reactor types may be accommodated. Preferably, alternating sleeves 24 are provided within angular-shaped separator 26, best seen in FIG. 3. Separators 26 are preferably secured to each of the four corners of a sleeve by welding separator 26 to each of the four corner of sleeve 24. Separators 26 provide a means to maintain a uniform space between adjacent sleeves. Positioned between separators 26 are sheets of neutron absorbing material 28 which serve to maintain fission reactions within basket 10 below a critical level necessary to sustain a fission chain reaction. The sheets of neutron absorbing material 28 are positioned and secured along the sides of each sleeve 24 by fastening means such as thin strips of steel 30 or other durable, resilient material such as steel alloy located intermittently along the length of the sleeve. The sheets of neutron absorbing material may comprise materials such as boron-carbide, aluminum powder, aluminum alloy, or the like. The steel strips 30 are preferably welded to separators 26 along each edge of the sleeve to hold the sheet of neutron absorbing material 28 in position. Referring now to FIG. 2, independent sleeves 24 are preferably positioned and held in place within basket 10 by a support element means preferably comprising a support structure with two separate plates 32 and 34 preferably composed of steel, steel alloy, or other durable resilient material. Plates 32 and 34 are positioned in and fill a gap between the inner wall 20 of cylindrical shell 12 and the perimeter of sleeve assembly 22. As seen in FIG. 2, plates 32 and 34 are preferably installed at multiple locations around the inner perimeter of basket 12. Plates 32 and 34 bear against the sleeves 24 and the inner wall 20 of cylindrical shell 12, however, they are preferably not attached by any fastening means to either. In FIG. 1 a bottom plate 36 is shown and is preferably welded to cylindrical shell 12 providing vertical support means for sleeves 24 and support plates 32 and 34 best seen in FIG. 4. Bottom plate 36 is preferably composed of a durable, resilient, non-corrosive material such as steel, steel alloy, or the like, and may be secured to cylindrical shell 12 by welds or other mechanical fastening means. Referring now to FIGS. 1 and 5, a shield lid 38 and structural lid 40 are shown installed on basket 10. Shield lid 38 provides shielding from radiation emanating from fuel assemblies contained in sleeves 24. Shield lid 38 is preferably composed of a plurality of steel disks 39 welded together and which preferably sandwich a section of the sheet of neutron absorbing material 28. Structural lid 40 is preferably a thick steel disk configured for attachment of hoist rings used to lift basket 10 after it has been loaded. Both shield lid 38 and structural lid 40 are preferably welded to cylindrical shell 12 and have access means, preferably penetrations 42, best seen in FIG. 5, for draining basket 10, vacuum drying basket 10, and backfilling basket 10 with helium after shield lid 38 and structural lid 40 are installed. Penetrations 42 may be apertures or bores and are preferably sealed using multiple welds once the helium backfill process has been completed. Shield lid 38 is preferably supported during its installation by a shield support ring 44. In operation and use basket 10 is extremely versatile, reliable, and may accommodate a large number of boiling water fuel assemblies, preferably sixty-one, while meeting the stringent requirements established by regulatory authorities both in the United States and abroad to ensure safety during the storage or transportation of fuel assemblies. Basket 10, when contained within a cask, is designed to withstand a wide variety of environmental hazards including earthquakes, floods, tornadoes, and various other accidents such as vertical drops on unyielding surfaces and the like. The basket shell, lid, and supporting structures are such that forces imposed on the contained fuel assemblies 48 during such hazardous conditions or accidents are maintained below those that would cause failure of the basket. Cylindrical shell 12 with welded end plates 36 and lids 38 and 40 provide ample support to sleeves 24 during and shock, accident or other stresses, thereby preventing distortion and maintaining stresses in the sleeves within acceptable limits. Basket 10 may be subjected to temperatures which vary across the basket internals or temperature gradients. The unique configuration of basket 10 and its internal supports provide the basket components with the capability to withstand the effects of various forces imposed on the basket, such as those from a drop event, without constraining the basket such that temperature gradients cause additional stresses in the basket components. Basket 10 is configured to adequately dissipate heat generated by contained fuel assemblies 48. Basket 10 maintains temperature in the fuel assembly region below the level at which long term degradation of the assemblies could occur. Basket 10 provides a means to maintain fission reactions within the basket at a level which is significantly below the critical level necessary to sustain a fission chain reaction. This is achieved through the use of the sheet of neutron absorbing material 28 operably positioned between adjacent sleeves 24 in basket 10. Basket 10 is specifically designed and constructed to minimize radiation exposure to plant workers and to the general public when the basket is loaded with fuel assemblies and is contained within a transportation, shipping, or storage cask. As is evident from the above description, basket 10 may be provided composed of a variety materials used to construct various parts of the basket without jeopardizing or limiting the ability of the basket to meet the applicable regulatory criteria. For example, cylindrical shell 12 may be constructed of carbon steel, stainless steel, or other metallic alloys. Sleeves 24 may be composed, for example, of carbon steel, stainless steel, or other metallic alloys. Additional advantages and modification will readily occur to those skilled in the art. The invention in its broader aspects is, therefore, not limited to the specific details, representative apparatus and illustrative examples shown and described. Accordingly, departures from such details may be made without departing from the spirit or scope of the applicant's general inventive concept.
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