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	abstract	A method for generating extreme ultraviolet radiation and soft x-ray radiation with a gas discharge operated on the left branch of the Paschen curve, in particular, for EUV lithography,
	summary
055747663	summary	FIELD OF THE INVENTION The invention relates to x-ray beam limiters. It finds particular application in conjunction with x-ray sources for computed tomographic ("CT") scanners and will be described with particular reference thereto. It will be appreciated, however, that the invention may find further application in other areas where resistance to contamination, ease of manufacture, and physical size constraints are important in the design or selection of a bearing for such devices. DESCRIPTION OF THE RELATED ART In certain x-ray equipment, including x-ray equipment used in computed tomography applications, it is desirable to change the fan angle of the x-ray beam traversing a scan area. An apparatus and method for performing this function is described in U.S. Pat. No. 4,905,268 entitled Adjustable Off-Focal Aperture for X-Ray Tubes to Mattson, et al., incorporated by reference herein. As more fully described in Mattson, a beam limiting plate is rotatably mounted within the path of the x-ray beam such that the position of the plate and the corresponding fan beam width may be changed by rotating the plate. With particular reference to FIG. 3 of Mattson, the rotatable mounting was accomplished using ball bearings. Although ball bearing beam limiting devices have functioned acceptably, aspects of these devices may be improved. Particulate matter may become deposited between the ball bearings, causing the balls to become jammed, thus causing the device to stick. The device must be serviced to alleviate this condition. Various attempts have been made to alleviate the effects of such particulate contamination. One approach was to remove one of the ball bearings from the assembly to increase clearance, but this is at best a temporary solution. More recently, various bearing platings were used in order to eliminate the use of dry bearing lubricant. This approach ultimately did not prevent the accumulation of particles between the balls. Yet another approach was to reduce the entry of contaminants by surrounding the assembly with a physical barrier. This approach proved unsatisfactory, however, primarily because particulates from sources internal to the assembly can still contribute to contamination. Another disadvantage associated with the use of prior art bearings is the extensive labor required in the manufacturing process. In the prior art, approximately sixty balls were used within each bearing. Accordingly, the use of bearings other than ball bearings would simplify the manufacturing process. One alternative approach which eliminates the need for ball bearings is the use of conventional sleeve or cylinder bearings. Conventional techniques, however, require the use of one or more snap rings to hold the bearing in place. The snap rings require additional physical space within the bearing assembly and may also require additional machining to create grooves for the snap rings. For the foregoing reasons, there is a need for an x-ray beam limiting device which is exhibits increased reliability, is simpler to manufacture, and which requires a minimum of physical space and componentry. SUMMARY According to one aspect of the present invention, a sleeve bearing for use with an x-ray beam limiting device is provided. The sleeve bearing includes a disc characterized by a diameter and a radiused outer edge, the outer edge having a bearing surface. The bearing further comprises means for permitting the diameter of the disc to be reduced to a compressed diameter when a compressive force is applied to the disc and for urging the disc to return to substantially its original diameter when the compressive force is removed. The sleeve bearing also includes a bearing retaining race which has an inner diameter smaller than the uncompressed diameter of the disc. There is a circular groove within the inner diameter of the race. The groove has a diameter and radius sized to accept the disc and permit the disc to rotate. Thus, the disc can be inserted in the groove when a compressive force is applied to the disc. The disc is rotatably retained within the groove when the compressive force is removed. It will be appreciated that the foregoing eliminates the need for ball bearings as well as retaining rings associated with conventional sleeve bearings. It will further be appreciated that the invention obviates the need for bearing lubrication, reduces susceptibility to particulate contamination, and simplifies assembly. According to another aspect of the invention, the bearing surface comprises polytetrafluoroethylene, and the bearing race is stainless steel. In yet other aspects of the present invention the disc is made from a resin containing fiberglass or from a polyetherimide resin. In yet another aspect of the invention, the beam limiter further includes an actuator means for causing a predetermined angular rotation of the disc, spring means for urging the disc to return to a predetermined position, and an x-ray beam limiter mounted to the disc. In another aspect of the invention, the means which permit the diameter of the disc to vary when a compressive force is removed is an aperture within the disc and a slot extending from the aperture to the outer edge of the disc. According to another aspect of the present invention, the sleeve bearing retaining race has an inner diameter with a circular groove characterized by a diameter and a radius disposed therein. The sleeve bearing further includes a disc having a radiused outer edge, an aperture, and a slot extending between the aperture and the outer edge. The outer edge of the disc has a bearing surface, and the disc with the bearing surface applied to it has an uncompressed diameter which is less than or equal to the diameter of the groove and which is retained within the groove. The disc is thereby rotatably mounted within the retaining race when the compressive force is removed. In another aspect of the invention, the beam limiter comprises a bearing retaining race having an inner diameter and a groove disposed within the inner diameter. A rotating assembly includes a disc having a bearing surface applied to its outer diameter and which is retained within the groove so that the disc is rotatable about an axis orthogonal to the plane of the disc. A radiation attenuating means having at least first and second differently dimensioned radiation passing positions which cross at the axis of rotation is mounted to the disc. The disc further contains means for rotating the disc between at least first and second positions.
	description	The present invention relates to a neutron shielding material composition. Further, the present invention relates to a neutron shielding material composition that is a material applied to a cask as a container for storing and transporting a spent nuclear fuel, exhibits improved heat resistance and has ensured neutron shielding performance. Nuclear fuels spent in nuclear facilities such as nuclear power plants are typically transported to reprocessing plants and then reprocessed. However, such spent nuclear fuels today are generated in an amount exceeding the reprocessing capacity. Thus, it is necessary to store spent nuclear fuels for a long period. In this case, spent nuclear fuels are cooled to a radioactivity level that makes the fuels suitable for transportation, and then placed in a cask as a nuclear shielding container and transported. Even at this stage, the spent nuclear fuels still emit radiation such as neutrons. Neutrons have high energy, and generate γ-rays to cause serious harm to the human body. For this reason, it is necessary to develop a material that surely shields such neutrons. Neutrons are known to be absorbed by boron. To make boron absorb neutrons, it is necessary to slow down the neutrons. Hydrogen is known to be most suitable as a substance for slowing down neutrons. Accordingly, a neutron shielding material composition must contain a large amount of boron and hydrogen atoms. Further, since spent nuclear fuels or the like as a neutron source generate decay heat, the fuels are heated to a high temperature when sealed for transportation or storage. Although the highest temperature varies depending upon the types of spent nuclear fuels, it is said that the temperature of spent nuclear fuels for high burnup may reach about 200° C. in a cask. For this reason, a nuclear shielding material for use preferably endures under such high-temperature conditions for about 60 years as a reference storage period for spent nuclear fuels. In this situation, use of a substance having a high hydrogen density, in particular, water as a shielding material has been proposed, and some of the proposals have been put into practice. However, water is difficult to be handled because it is a liquid, and is not suitable for a cask for transportation and storage, in particular. Moreover, it is difficult to suppress boiling in a cask in which the temperature reaches 100° C. or more, disadvantageously. Conventionally, a resin composition has been used as a material for a neutron shielding material, and an epoxy resin has been used in one of such resin compositions. Generally, there is a reciprocal relationship between hydrogen content and heat resistance in a resin composition. A resin composition having a high hydrogen content tends to have low heat resistance, and a resin composition having high heat resistance tends to have a low hydrogen content. An epoxy resin exhibits excellent heat resistance and curability, but tends to contain only a small amount of hydrogen indispensable for slowing down neutrons. Therefore, an amine curing agent having a high hydrogen content has been used to compensate this drawback. Japanese Patent Laid-Open No. 6-148388 discloses a neutron shielding material composition which employs a polyfunctional amine epoxy resin to have reduced viscosity and improved workability at ordinary temperature and exhibits excellent pot life. Japanese Patent Laid-Open No. 9-176496 discloses a neutron shielding material obtained by curing a composition made of an acrylic resin, epoxy resin, silicone resin or the like with a polyamine curing agent. Since an amine compound has a relatively high hydrogen content, the effect of absorbing neutrons is improved. However, the carbon-nitrogen bond contained in an amine curing agent is easily decomposed by heat. Accordingly, it has been demanded to develop a novel composition having durability necessary for storing a spent nuclear fuel for high burnup, rather than a conventional neutron shielding material made of a resin cured with an amine curing agent. An object of the present invention is to provide a neutron shielding material composition which exhibits thermal durability improved as compared with a conventional composition, and surely absorbs neutrons. The present invention provides a neutron shielding material composition comprising a polymerization initiator, a polymerization component, a density increasing agent and a boron compound. The present invention provides a neutron shielding material composition not comprising a curing agent. The composition preferably comprises an epoxy component as the polymerization component. The composition particularly preferably comprises a hydrogenated epoxy compound as the epoxy component. The hydrogenated epoxy compound herein refers to an epoxy compound having an increased hydrogen content obtained by hydrogenating at least part of a benzene ring to break conjugation of the part of the benzene ring but nevertheless maintain the cyclic structure. In the present invention, the epoxy component preferably comprises a compound of the structural formula (1): wherein X is at least one compound selected from compounds of the structural formulas (2), (3), (4), (5) and (6): wherein R1 to R4 are each independently selected from the group consisting of CH3, H, F, Cl and Br, and n is 0 to 2 in the structural formula (2), R5 to R8 are each independently selected from the group consisting of CH3, H, F, Cl and Br, and n is 0 to 2 in the structural formula (3), n is 1 to 12 in the structural formula (5), and n is 1 to 24 in the structural formula (6). The epoxy component preferably comprises a compound of the structural formula (14): wherein n is 1 to 3. The epoxy component also preferably comprises at least one compound selected from the group consisting of compounds of the structural formulas (7), (8), (15), and (17): wherein R9 is a C1-10 alkyl group or H, and n is 1 to 24; a compound of the structural formula (8): wherein n is 1 to 8; a compound of the structural formula (15) wherein n is 1 to 3; and a compound of the structural formula (17). The neutron shielding material composition of the present invention preferably further comprises a compound for increasing the hydrogen content of the composition. The composition preferably comprises, as the compound for increasing the hydrogen content, at least one of compounds of the structural formulas (9) and (10): wherein n is 1 to 3. The composition preferably comprises an oxetane compound as the polymerization component, and the oxetane compound preferably comprises at least one of compounds of the structural formulas (19) and (20). Further, the polymerization initiator preferably comprises a cationic polymerization initiator, and the cationic polymerization initiator preferably comprises a compound of the structural formula (11) or (16): wherein R10 is a hydrogen atom, a halogen atom, a nitro group or a methyl group, R11 is a hydrogen atom, CH3CO or CH3OCO, and X is SbF6, PF6, BF4 or AsF6. The density increasing agent is preferably a metal powder having a density of 5.0 to 22.5 g/cm3, a metal oxide powder having a density of 5.0 to 22.5 g/cm3, or a combination thereof. The neutron shielding material composition of the present invention preferably further comprises a filler, and preferably further comprises a refractory material. The refractory material preferably comprises at least one of magnesium hydroxide and aluminum hydroxide. Magnesium hydroxide is more preferably magnesium hydroxide obtained from sea water magnesium. The present invention further provides a neutron shielding material and a neutron shielding container produced from the neutron shielding material composition. Reaction in the composition of the present invention proceeds between a compound polymerizable by the action of a polymerization initiator, preferably an epoxy component, and a polymerization initiator, and the composition does not comprise an amine curing agent susceptible to heat. Thus, a cask using the composition of the present invention as a material has improved heat resistance. The composition also has a hydrogen content satisfying the standard, and has ensured neutron shielding performance. Further, since the composition of the present invention comprises a density increasing agent, the neutron shielding material can provide an increased neutron absorption while maintaining secondary γ-ray shielding performance, and accordingly can have improved neutron shielding performance without placing a structure for shielding γ-rays outside the main body of the neutron shielding material as in a conventional manner. Embodiments of the present invention will be described in detail below. The embodiments described below do not limit the present invention. Throughout the present invention, a polymerization component refers to a compound polymerizable by the action of a polymerization initiator. In particular, the composition of the present invention comprises, as polymerization components, an epoxy component and an oxetane component described below. An epoxy component refers to a compound having an epoxy ring (herein after referred to as epoxy compound), and may be one epoxy compound or a mixture of two or more epoxy compounds. Similarly, an oxetane compound refers to a compound having an oxetane ring, and may be one oxetane compound or a mixture of two or more oxetane compounds. A resin component refers to a combination of a polymerization component as described above with a polymerization initiator component, and a combination of these components with a compound for increasing the hydrogen content, for example, a diol. In the present invention, the composition can be cured without using a curing agent having an amine moiety susceptible to heat by adding a polymerization initiator component to a cationically polymerizable compound, in particular, an epoxy compound, an oxetane compound or both. A conventional composition employs an amine compound as a curing agent, and thus has decreased heat resistance, in particular, thermal decomposition resistance in a high-temperature condition for a long period. Since the composition of the present invention can be cured without use of such a curing agent, a resin having no carbon-nitrogen bond moiety in which the bond is easily decomposed in a high-temperature state can be obtained, and high heat resistance can be expected. Accordingly, since a decrease in heat resistance by use of a curing agent does not occur as in a conventional composition, the composition of the present invention can be provided with desired properties such as hydrogen content and heat resistance by selection of a polymerization component. The composition of the present invention is preferably a composition having a high hydrogen content comprising a polymerization component, a polymerization initiator component, a density increasing agent, a boron compound as a neutron absorbent, and a refractory material, characterized in that the composition is cured to be a resin with high heat resistance and high neutron shielding effect. Specifically, the composition of the present invention is required to have a temperature of 330° C. or more, and preferably 350° C. or more for attaining a residual weight ratio of 90 wt % by thermogravimetric analysis of a cured product thereof, and to have a hydrogen content of preferably 9.0 wt % or more, and more preferably 9.8 wt % or more based on the total resin component. This is because, if the hydrogen content is 9.0 wt % or more, neutron shielding effect to be achieved can be ensured by controlling the amount of the refractory material. In addition, more specifically, the cured product after thermal endurance in a high-temperature closed environment for a long period preferably has a weight reduction and compressive strength as small as possible. For example, the cured resin after thermal endurance in a closed environment at 190° C. for 1,000 hours is required to have a weight reduction of 0.5 wt % or less, and preferably 0.2 wt % or less, and to have compressive strength not reduced, and most preferably inclined to be increased instead. As the polymerization component of the present invention, a compound having high heat resistance is preferably used. An epoxy compound is particularly preferably used, since the composition requires heat resistance at 100° C. or more, and preferably at about 200° C. As the epoxy component of the present invention, a compound having an epoxy ring which can be polymerized using a cationic polymerization initiator component is used. To improve heat resistance, the epoxy component preferably has a high crosslinking density. In addition, when the epoxy component contains many ring structures, the compound has a rigid structure, and thus can improve heat resistance. Examples of the ring structure include a benzene ring. A benzene ring is rigid and has excellent heat resistance, but contains only a small amount of hydrogen that functions to slow down neutrons in the present invention. Thus, a compound with a hydrogenated benzene ring is more preferable. As a rigid structure having high heat resistance, a structure represented by the formula (12) is preferable. A structure represented by the formula (13) is most preferable, because such a rigid structure preferably has a higher hydrogen content. Throughout the present specification, such an epoxy compound having a ring structure in which a benzene ring is hydrogenated is referred to as a hydrogenated epoxy compound. A hydrogenated epoxy compound has a heat-resistant structure and a high hydrogen content, and is thus most preferable as the epoxy compound of the present invention. The epoxy component may be one epoxy compound or a mixture of a plurality of epoxy compounds. An epoxy compound is selected so that the compound can impart desired properties such as increased heat resistance and hydrogen content. The composition of the epoxy component is determined so that the resin component contains hydrogen in an amount sufficient for shielding neutrons, and preferably in an amount of preferably 9.0 wt % or more, and more preferably 9.8 wt % or more. Neutron shielding performance of the neutron shielding material is determined according to the hydrogen content (density) of the neutron shielding material and the thickness of the neutron shielding material. This value is based on the hydrogen content required for the resin component, which is calculated with respect to the hydrogen content (density) required for the neutron shielding material, determined from neutron shielding performance required for a cask and the designed thickness of the neutron shielding material in the cask, taking into consideration the amounts of the refractory material and the neutron absorbent added to the neutron shielding material and kneaded. From this point of view, a compound having an epoxy ring, preferably a plurality of epoxy rings, which has a rigid structure or a ring structure represented by the structural formula (12) or (13) and has a high hydrogen content is suitable as the epoxy component of the present invention. Such an epoxy component is generally represented by the structural formula (1), wherein X is preferably selected from the structural formula (2), wherein R1 to R4 are each independently selected from the group consisting of CH3, H, F, Cl and Br, and n is 0 to 2, the structural formula (3), wherein R5 to R8 are each independently selected from the group consisting of CH3, H, F, Cl and Br, and n is 0 to 2, the structural formula (4) or (5), wherein n is 1 to 12, and the structural formula (6), wherein n is 1 to 24. In particular, a hydrogenated bisphenol A epoxy represented by the structural formula (14) is used as a most suitable and important epoxy component to provide a hydrogen content and heat resistance in a well-balanced manner. Further, a bisphenol A epoxy (structural formula (15)) may be added as a component for imparting heat resistance. This is because the compound has a benzene ring and a rigid structure. To increase crosslinking density and improve heat resistance, the structural formula (7), wherein R9 is a C1-10 alkyl group or H, and n is 1 to 24, the structural formula (8), wherein n is 1 to 8, or the structural formula (19) is preferably added. Accordingly, a mixture of the structural formula (14) with at least one compound selected from the group consisting of the structural formula (15), the structural formula (7), the structural formula (8) and the structural formula (17) can provide a compound having desired hydrogen content and heat resistance. Thus, the epoxy component of the present invention comprises an epoxy compound represented by the structural formula (14), and may comprise all or some of the structural formula (15), the structural formula (7), the structural formula (8) and the structural formula (17). Any possible combination using these epoxy compounds can be used. In this case, the composition preferably comprises 70 wt % or more of a hydrogenated bisphenol A epoxy of the structural formula (14), 20 wt % or less of a bisphenol A epoxy of the structural formula (15), 30 wt % or less of the structural formula (7), 25 wt % or less of the structural formula (8) and 30 wt % or less of the structural formula (17), respectively based on the total resin content. In particular, an oxetane compound can be used as the polymerization component to increase the hydrogen content. An oxetane compound can be cationically polymerized like an epoxy, has a high hydrogen content, and is expected to have certain heat resistance. Generally, an oxetane compound is represented by the structural formula (18): wherein R12 and R13 are each independently H, halogen, C1-8 alkyl, an alcohol, or another structure containing an organic compound composed of carbon, hydrogen and oxygen. The oxetane compound used in the present invention may be a compound having two or more oxetane rings through an ether bond or benzene ring. Specifically, the oxetane compound used in the present invention is preferably the structural formula (19) or the structural formula (20). The oxetane compound is not limited thereto. A compound having at least two oxetane rings through, for example, an ether bond or ring structure like the structural formula (19) is preferable. This is because a compound containing many oxetane rings can be expected to impart heat resistance by increasing the crosslinking density. Further, an oxetane compound having many ring structures, branched structures or the like is preferable, since the composition of the present invention is particularly required to be provided with heat resistance. An oxetane component may be used singly as the polymerization component without using an epoxy compound. Two or more oxetane compounds may be used. An oxetane component may be used as the polymerization component in combination with any epoxy component. Preferable examples of a combination of polymerization components include a combination of an oxetane component of the structural formula (19) with an epoxy component of the structural formula (7), a combination of an oxetane component of the structural formula (19) with an epoxy component of the structural formula (8), and a combination of an oxetane component of the structural formula (19) with an epoxy component of the structural formula (17). In one example of a composition ratio of polymerization components using an oxetane compound, the structural formula (19) is 85.5 wt % and the structural formula (15) is 14.5 wt %. In another example, the structural formula (19) is 74.0 wt %, the structural formula (20) is 20.0 wt %, and the structural formula (7) is 6.0%. Polymerization initiators are classified into radical polymerization initiators, anionic polymerization initiators, cationic polymerization initiators, and the like, and many of them are reported in documents or the like. In the present invention, cationic polymerization initiators are preferably used. Examples of well-known cationic polymerization initiators are shown in Table 1. Examples of cationic thermal polymerization initiators that can initiate polymerization by heat include Opton CP series of Asahi Denka Co., Ltd.; SI series of Sanshin Chemical Industry Co., Ltd.; and DAICAT EX-1 of Daicel Chemical Industries, Ltd. These polymerization initiators can be used, but are not exclusively used, in the present invention. TABLE 1General polymerization initiator componentsProductStructurenameSupplierX = SbF6 UVI-6974UCC X = PF6 UVI-6990UCC X = SbF6 UVI-6970 (SP-170)X = PF6 UVI-6950 (SP-150)Asahi Denka Degacure K126 FX-512Degussa 3M X = SbF6 PIC-061TX = PF6 PIC-062TNippon Kayaku X = SbF6 PIC-020TX = PF6 PIC-022TNippon Kayaku Synthetic sulfonium salt Nippon Soda UV-9380CGE IOC-10GE CD-1012Sartomer 2074Rhone-Poulenc Chimie Iruga-cure 261Chiba-Geigy Toshiba As the polymerization initiator, a compound represented by the structural formula (11) or (16) is preferably added. The polymerization initiator is added in an amount of preferably 0.5 to 6 parts by weight, and more preferably 1 to 3 parts by weight based on 100 parts by weight of the total resin component. This is because, if the polymerization initiator is added too much, the hydrogen content in the total composition may be decreased. Further, a compound that does not have an epoxy ring and contains a large amount of hydrogen may be added to the composition of the present invention to increase the hydrogen content. Such a compound may be optionally added when the hydrogen content is insufficient, since the hydrogen content cannot be indefinitely increased by an epoxy compound alone. Here, the compound to be added must be selected so that the compound does not significantly affect properties of the entire system of the composition. For example, when an amine compound is mixed with the composition of the present invention containing a cationic polymerization initiator, polymerization reaction of the epoxy component does not proceed. Therefore, an amine compound cannot be added. As a result of studies taking this point into consideration, a diol is suitable as a compound for increasing the hydrogen content, for example. Any diol can be used insofar as it is soluble in the epoxy component and polymerizable with the epoxy component. Examples of the diol that can be used include, but are not limited to, an aliphatic diol, an aromatic diol, and a diol or polyol having an alicyclic structure. Preferably, a diol having an alicyclic structure, for example, a compound represented by the structural formula (9) or (10) is used in order to increase the hydrogen content and suppress a decrease in heat resistance. A diol is added in an amount of preferably 30 wt % or less, and more preferably 20 wt % or less based on the total resin component. The compound for increasing the hydrogen content in the composition is not limited to a diol. A cationically curable oxetane or vinyl ether, a trifunctional or higher functional alcohol that can expected to have the same effect as in a diol, or the like can be used. The density increasing agent may be any material that is dense and can increase the specific gravity of the neutron shield, unless the material adversely affects other components. Here, the density increasing agent itself which effectively shields γ-rays has a density of 5.0 g/cm3 or more, preferably 5.0 to 22.5 g/cm3, and more preferably 6.0 to 15 g/cm3. If the density is 5.0 g/cm3 or less, it is difficult to effectively shields γ-rays without impairing neutron shielding capability. If the density is 22.5 g/cm3 or more, an effect in proportion to the amount added cannot be observed. Specific examples of the density increasing agent include metal powders and metal oxide powders. Preferable examples of the density increasing agent include metals having a melting point of 350° C. or more such as Cr, Mn, Fe, Ni, Cu, Sb, Bi, U and W; and metal oxides having a melting point of 1,000° C. or more such as NiO, CuO, ZnO, ZrO2, SnO, SnO2, WO2, UO2, PbO, WO3 and lanthanoid oxides. Of these, Cu, WO2, WO3, ZrO2 and CeO2 are particularly preferable. This is because they are advantageous in terms of cost. The density increasing agent may be used singly or in a mixture of two or more. There are no specific limitations to the particle size of the density increasing agent. However, if the particle size is large, the density increasing agent may settle during production. Therefore, the particle size is preferably small to the extent that settling does not occur. The particle size that does not cause settling largely depends on other conditions (for example, the temperature, viscosity, curing speed and the like of the composition), and thus cannot be numerically defined simply. By adding the density increasing agent, the specific gravity of the neutron shield can be increased, and γ-rays can be more effectively shielded. By use of the above-described metal powder or metal oxide powder, fire resistance can also be improved. By replacing a part of an additive other than the resin component, mainly a part of the refractory material with the density increasing agent, the hydrogen content may be increased. By replacing mainly a part of the refractory material with the density increasing agent, the amount of the epoxy resin can be increased while maintaining the specific gravity of the neutron shielding material composition (1.62 to 1.72 g/cm3). Thus, a neutron shield having a high hydrogen content can be produced, and neutrons can be effectively shielded. Specifically, neutron shielding capability and γ-ray shielding can be achieved at the same time. The amount of the density increasing agent to be added can be appropriately adjusted to maintain the specific gravity of the above-described neutron shielding material composition (1.62 to 1.72 g/cm3). It is difficult to specifically define the amount, because the amount varies according to the type of the density increasing agent used, the types and contents of other components, and the like. For example, the amount is 5 to 40 mass %, and preferably 9 to 35 mass % based on the total neutron shielding material composition. The amount is particularly preferably 15 to 20 mass % when using CeO2. If the amount is 5 mass % or less, it is difficult to observe the effect of adding the density increasing agent. If the amount is 40 mass % or more, it is difficult to maintain the specific gravity of the neutron shielding material composition at 1.62 to 1.72 g/cm3. Examples of a boron compound used as the neutron absorbent in the composition of the present invention include boron carbide, boron nitride, boric acid anhydride, boron iron, colemanite, orthoboric acid and metaboric acid. Boron carbide is most preferable in terms of neutron shielding performance. The above-described boron compound is used as a powder without specific limitations to its particle size and amount added. However, taking dispersibility in the epoxy resin of the matrix resin and neutron shielding performance into consideration, the average particle size is preferably about 1 to 200 microns, more preferably about 10 to 100 microns, and particularly preferably about 20 to 50 microns. On the other hand, the amount of the boron compound added is most preferably 0.5 to 20 wt % based on the total composition including the filler described below. If the amount is less than 0.5 wt %, the boron compound added exhibits only a small effect as the neutron shielding material. If the amount is more than 20 wt %, it is difficult to homogeneously disperse the boron compound. In the present invention, a powder of silica, alumina, calcium carbonate, antimony trioxide, titanium oxide, asbestos, clay, mica or the like; a glass fiber; or the like is used as the filler. A carbon fiber or the like may be added if necessary. Further, if necessary, a natural wax, fatty acid metal salt, acid amide, fatty acid ester or the like as a releasing agent; paraffin chloride, bromotoluene, hexabromobenzene, antimony trioxide or the like as a flame retardant; carbon black, iron oxide red or the like as a colorant; a silane coupling agent; a titanium coupling agent; or the like can be added. The refractory material used in the composition of the present invention aims to preserve a certain amount or more of the neutron shielding material so that neutron shielding capability can be maintained to a certain extent or higher even in case of fire. As such a refractory material, magnesium hydroxide or aluminum hydroxide is particularly preferable. Of these, magnesium hydroxide is particularly preferable, because it is present in a stable manner even at a high temperature of about 200° C. Magnesium hydroxide is preferably magnesium hydroxide obtained from sea water magnesium. This is because magnesium in sea water has a high purity to make the hydrogen ratio in the composition relatively high. Sea water magnesium can be produced by a method such as a sea water method or ionic brine method. Otherwise, a commercially available product Kisuma 2SJ (product name, Kyowa Chemical Industry Co., Ltd.) may be purchased and used. However, commercially available magnesium hydroxide is not limited to this product. The refractory material is added in an amount of preferably 20 to 70 wt %, and particularly preferably 35 to 60 wt % based on the total composition. The composition of the present invention is prepared by mixing a polymerization component, for example, an epoxy component with other additives to prepare a resin composition; kneading the resin composition with a refractory material, a neutron absorbent or the like; and finally adding a polymerization initiator. Although polymerization conditions differ according to the composition of the resin component, heating is preferably carried out at a temperature of 50° C. to 200° C. four 1 to 3 hours. Further, such heating treatment is preferably carried out in two stages. It is preferable to carry out heating treatment at 80° C. to 120° C. for 1 to 2 hours, and then at 120° C. to 180° C. four 2 to 3 hours. However, the preparation method, curing conditions and the like are not limited thereto. Further, a container, preferably a cask, for effectively shielding neutrons in a spent nuclear fuel and storing and transporting the spent nuclear fuel can be produced. Such a transportation cask can be produced utilizing a known technology. For example, in a cask disclosed in Japanese Patent Laid-Open No. 2000-9890, a location to be filled with a neutron shield is provided. Such a location can be filled with the composition of the present invention. The composition of the present invention can be used not only for such a shield, but also for various places in apparatuses and facilities to prevent diffusion of neutrons, and can effectively shield neutrons. Specific examples of embodiments of the present invention using a resin component, a density increasing agent and a refractory material will be further described in detail with reference with the drawings. Here, embodiments in which a boron compound or a filler is not added will be described for illustration. However, the present invention is not limited to such embodiments. FIG. 1 is a conceptual view showing a configuration example of the neutron shield of the present embodiment. Specifically, as shown in FIG. 1, the neutron shield of the present embodiment is obtained by mixing a resin component 1 comprising a polymerization component and a polymerization initiator as main components with a refractory material 2 and a density increasing agent 3 having a density higher than in the refractory material 2. Here, the neutron shield is provided with an increased hydrogen content while maintaining the material density (in the range of 1.62 to 1.72 g/mL), by mixing a metal powder or metal oxide powder as the density increasing agent 3, in particular. The density increasing agent 3 to be mixed has a density of 5.0 g/mL or more, preferably 5.0 to 22.5 g/mL, and more preferably 6.0 to 15 g/mL. Further, the density increasing agent 3 to be mixed is preferably a metal powder having a melting point of 350° C. or more or a metal oxide powder having a melting point of 1,000° C. or more. Examples of a powder material corresponding to the density increasing agent include metals such as Cr, Mn, Fe, Ni, Cu, Sb, Bi, U and W. Further examples thereof include metal oxides such as NiO, CuO, ZnO, ZrO2, SnO, SnO2, WO2, CeO2, UO2, PbO, PbO, and WO3. Since the neutron shield of the present embodiment configured as above is prepared by mixing the resin component 1 comprising a polymer as a main component, the refractory material 2, and the density increasing agent 3 having a density higher than in the refractory material 2, the neutron shield can have an increased hydrogen content while maintaining the material density at a certain value (in the range of 1.62 to 1.72 g/mL). Specifically, the refractory material 2 has a slightly higher density and a slightly lower hydrogen content as compared with the neutron shielding material 1. Thus, a part of the refractory material 2 is replaced with the density increasing agent 3 not containing hydrogen to make the material density equal. By calculating the density and the hydrogen content of each component and carrying out appropriate replacement, the refractory material 2 having a slightly lower hydrogen content is replaced with the resin component 1 having a high hydrogen content, so that the neutron shield can have an increased hydrogen content. As a result, the neutron shield can provide an increased neutron absorption while maintaining secondary γ-ray shielding performance, and accordingly can have improved neutron shielding performance without placing a structure for shielding γ-rays outside the main body of the neutron shield as in a conventional manner. In the neutron shield of the present embodiment, the density increasing agent 3 to be mixed has a density of 5.0 g/mL or more, preferably 5.0 to 22.5 g/mL, and more preferably 6.0 to 15 g/mL. Therefore, the neutron shield can exhibit the above-described effect more significantly. FIG. 2 is a characteristic view showing the relation between the density of the density increasing agent 3 and the hydrogen content. FIG. 2 shows a hydrogen content of the neutron shield originally having a hydrogen content of 0.0969 g/mL, containing magnesium hydroxide as the refractory material 2 and containing the resin component 1 having a density of 1.64 g/mL, in which the refractory material 2 is replaced with the density increasing agent 3 to make the material density constant. Magnesium hydroxide as the refractory material 2 has a density of 2.36 g/mL. As is clear from FIG. 2, the density increasing agent 3 is effective only if the density of the density increasing agent 3 reaches a density slightly higher than in the refractory material 2, not the density of the refractory material 2, although the effective density differs according to the resin component 1 and the refractory material 2. Specifically, the density increasing agent 3 is effective at a density of 5.0 g/mL or more, and preferably 6.0 g/mL or more. If the density is 22.5 g/mL or more, an effect in proportion to the amount added cannot be observed. FIG. 3 is a characteristic view showing the relation between the density of the density increasing agent 3 and the relative ratio of the neutron and secondary γ-ray dose outside the neutron shield. FIG. 3 shows a shielding effect of the neutron shield originally having a hydrogen content of 0.0969 g/mL, containing magnesium hydroxide as the refractory material 2 and containing the base resin 1 having a density of 1.64 g/mL, in which the refractory material 2 is replaced with the density increasing agent 3 to make the material density constant. The dose outside the shield of the resin component 1 is defined as “1”. As is clear from FIG. 3, the effect can be observed when the density increasing agent 3 has a density of 5.0 g/mL or more, and more preferably 6.0 g/mL or more. If the density is 22.5 g/mL or more, an effect in proportion to the amount added cannot be observed. Further, the neutron shield of the present embodiment can be provided with improved fire resistance by mixing a metal powder having a melting point of 350° C. or more (such as Cr, Mn, Fe, Ni, Cu, Sb, Bi, U or W) or a metal oxide powder having a melting point of 1,000° C. or more (such as NiO, CuO, ZnO, ZrO2, SnO, SnO2, WO2, CeO2, UO2, PbO, PbO or WO3). As described above, the neutron shield of the present embodiment can have an increased hydrogen content while maintaining the material density at a certain value without any decrease, and accordingly can have improved neutron shielding performance without placing a structure for shielding γ-rays outside the main body of the neutron shield as in a conventional manner. As shown in the above FIG. 1, the neutron shield of the present embodiment is obtained by mixing an epoxy component and a polymerization initiator as a resin component 1 with a refractory material 2 and a density increasing agent 3 having a density higher than in the refractory material 2, and forming the mixture by curing. The density increasing agent 3 to be mixed has a density of 5.0 g/mL or more, preferably 5.0 to 22.5 g/mL, and more preferably 6.0 to 15 g/mL. Further, the density increasing agent 3 to be mixed is preferably a metal powder having a melting point of 350° C. or more or a metal oxide powder having a melting point of 1,000° C. or more. Examples of a powder material corresponding to the density increasing agent include metals such as Cr, Mn, Fe, Ni, Cu, Sb, Bi, U and W. Further examples thereof include metal oxides such as NiO, CuO, ZnO, ZrO2, SnO, SnO2, WO2, CeO2, UO2, PbO, PbO, and WO3. Since the neutron shield of the present embodiment configured as above is prepared by mixing the resin component 1, the refractory material 2, and the density increasing agent 3 having a density higher than in the refractory material 2, the neutron shield can have an increased hydrogen content while maintaining the material density at a certain value (in the range of 1.62 to 1.72 g/mL). Specifically, the refractory material 2 has a slightly higher density and a slightly lower hydrogen content as compared with the resin component 1. Thus, apart of the refractory material 2 is replaced with the density increasing agent 3 not containing hydrogen to make the material density equal. By calculating the density and the hydrogen content of each component and carrying out appropriate replacement, the refractory material 2 having a slightly lower hydrogen content is replaced with the resin component 1 having a high hydrogen content, so that the neutron shield can have an increased hydrogen content. As a result, the neutron shield can provide an increased neutron absorption while maintaining secondary γ-ray shielding performance, and accordingly can have improved neutron shielding performance without placing a structure for shielding γ-rays outside the main body of the neutron shielding material as in a conventional manner. In the neutron shielding material of the present embodiment, the density increasing agent 3 to be mixed has a density of 5.0 g/mL or more, preferably 5.0 to 22.5 g/mL, and more preferably 6.0 to 15 g/mL. Therefore, the neutron shielding material can exhibit the above-described effect more significantly. FIG. 2 is a characteristic view showing the relation between the density of the density increasing agent 3 and the hydrogen content. FIG. 2 shows a hydrogen content of the neutron shield originally having a hydrogen content of 0.0969 g/mL, containing magnesium hydroxide as the refractory material 2 and containing the base resin 1 having a density of 1.64 g/mL, in which the refractory material 2 is replaced with the density increasing agent 3 to make the material density constant. Magnesium hydroxide as the refractory material 2 has a density of 2.36 g/mL. As is clear from FIG. 2, the density increasing agent 3 is effective only if the density of the density increasing agent 3 reaches a density slightly higher than in the refractory material 2, not the density of the refractory material 2, although the effective density differs according to the base resin 1 and the refractory material 2. Specifically, the density increasing agent 3 is effective at a density of 5.0 g/mL or more, and more preferably 6.0 g/mL or more. If the density is 22.5 g/mL or more, an effect in proportion to the amount added cannot be observed. FIG. 3 is a characteristic view showing the relation between the density of the density increasing agent 3 and the relative ratio of the neutron and secondary γ-ray dose outside the neutron shield. FIG. 3 shows a shielding effect of the neutron shield originally having a hydrogen content of 0.0969 g/mL, containing magnesium hydroxide as the refractory material 2 and containing the base resin 1 having a density of 1.64 g/mL, in which the refractory material 2 is replaced with the density increasing agent 3 to make the material density constant. The dose outside the shield of the base resin 1 is defined as “1”. As is clear from FIG. 3, the effect can be observed when the density increasing agent 3 has a density of 5.0 g/mL or more, and preferably 6.0 g/mL or more. If the density is 22.5 g/mL or more, an effect in proportion to the amount added cannot be observed. Further, the neutron shield of the present embodiment can be provided with improved fire resistance by mixing a metal powder having a melting point of 350° C. or more (such as Cr, Mn, Fe, Ni, Cu, Sb, Bi, U or W) or a metal oxide powder having a melting point of 1,000° C. or more (such as NiO, CuO, ZnO, ZrO2, SnO, SnO2, WO2, CeO2, UO2, PbO, PbO or WO3). As described above, the neutron shield of the present embodiment also can have an increased hydrogen content while maintaining the material density at a certain value without any decrease, and accordingly can have improved neutron shielding performance without placing a structure for shielding γ-rays outside the main body of the neutron shield as in a conventional manner. Specifically, since the neutron shield can be more effective for shielding neutrons while maintaining γ-ray shielding performance by use of a density increasing agent, it can be less necessary to place a heavy structure for shielding γ-rays outside the main body of the neutron shield as in a conventional manner. The present invention will be described in detail below by way of examples. The examples below do not limit the present invention. In the examples, the composition of the present invention was prepared, and the neutron shielding effect was examined. Typically, a resin composition for a neutron shielding material is mixed with copper as a density increasing agent, aluminum hydroxide or magnesium hydroxide as a refractory material, and a boron compound such as boron carbide as a neutron absorbent, respectively in an amount of about 20 wt %, about 40 wt % and about 1 wt % based on the total resin composition to prepare a neutron shield. However, compositions with a refractory material and a neutron absorbent not added are mainly described here in order to evaluate properties exhibited by a resin component, specifically, a polymerization component, a polymerization initiator component and the like, and a density increasing agent. Properties required for the neutron shielding material include heat resistance (residual weight ratio, compressive strength, or the like), fire resistance and hydrogen content (the material must have a certain hydrogen content density or higher in order to be judged suitable for a neutron shield). Since fire resistance largely depends upon the refractory material, the resin composition for a neutron shielding material was evaluated for its heat resistance represented by a residual weight ratio and hydrogen content. The residual weight ratio was determined by measuring the weight change during heating to evaluate heat resistance of the composition. TGA was used for the measurement. The weight reduction by heat was measured under a condition where the composition was heated from room temperature to 600° C. at a rate of temperature rise of 10° C./min in a nitrogen atmosphere. A hydrogen content in a single resin of 9.8 wt % or more was defined as the standard hydrogen content required for the resin. 1 g of a cationic polymerization initiator SI-80 (structural formula (11)) was added to 100 g of a hydrogenated bisphenol A epoxy resin (manufactured by Yuka Shell Epoxy K.K., YL6663 (structural formula (14)). The mixture was sufficiently stirred until the polymerization initiator was dissolved, and then mixed with 50 g of copper having a density of 8.92 g/cm3 as a density increasing agent to prepare a resin composition used for a neutron shielding material. As a result of measuring the hydrogen content in the resin composition for a neutron shielding material, the hydrogen content was 9.8 wt % or more (about 10 wt % or more) which satisfied the standard. Next, the composition was cured at 80° C. for 30 minutes and at 150° C. for 2 hours, and the weight reduction by heat of the cured product was measured by TGA. The weight reduction by heat was measured under a condition where the composition was heated from RT to 600° C. at a rate of temperature rise of 10° C./min in a nitrogen atmosphere. As a result of measurement, the residual weight ratio at 200° C. was 99.5 wt % or more, and the temperature at a residual weight ratio of 90 wt % was 350° C. or more, meaning that the composition exhibited extremely good heat resistance and heat stability. 1 g of a cationic polymerization initiator SI-80 (structural formula (11)) was added to a mixture of 84.6 g of a hydrogenated bisphenol A epoxy resin (YL6663, structural formula (14)) and 15.4 g of a bisphenol A epoxy resin (manufactured by Yuka Shell Epoxy K.K., Epicoat 828, structural formula (15)) as epoxy resins. The mixture was sufficiently stirred until the polymerization initiator was dissolved, and then mixed with 50 g of copper as a density increasing agent to prepare a resin composition used for a neutron shielding material. As a result of measuring the hydrogen content in the resin composition in the same manner as in Example 1, the hydrogen content was about 9.8 wt % which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was 99.5 wt % or more, and the temperature at a residual weight ratio of 90 wt % was 380° C. or more, meaning that the composition exhibited extremely good heat resistance and heat stability. 74.8 g of a hydrogenated bisphenol A epoxy resin (YL6663, structural formula (14)) and 25.2 g of a polyfunctional alicyclic epoxy resin (manufactured by Daicel Chemical Industries, Ltd., EHPE3150, structural formula (7)) were mixed as epoxy resins. The mixture was maintained at 110° C. and sufficiently stirred until EHPE3150 (solid) was dissolved. After dissolution of EHPE3150, the mixture was allowed to stand in an environment at room temperature. When the temperature of the mixture was lowered to about room temperature, 1 g of a cationic polymerization initiator SI-80 (structural formula (11)) was added, and the mixture was sufficiently stirred until the polymerization initiator was dissolved. 50 g of copper was mixed therewith as a density increasing agent to prepare a resin composition used for a neutron shielding material. As a result of measuring the hydrogen content in the resin composition, the hydrogen content was about 9.8 wt % which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was about 99.5 wt %, and the temperature at a residual weight ratio of 90 wt % was 390° C. or more, meaning that the composition exhibited extremely good heat resistance and heat stability. 1 g of a cationic polymerization initiator SI-80 (structural formula (11)) was added to a mixture of 79.4 g of a hydrogenated bisphenol A epoxy resin (YL6663, structural formula (14)) and 20.6 g of an alicyclic epoxy resin (manufactured by Daicel Chemical Industries, Ltd., Celloxide 2021P, structural formula (8)) as epoxy resins. The mixture was sufficiently stirred until the polymerization initiator was dissolved, and then mixed with 50 g of copper as a density increasing agent to prepare a resin composition used for a neutron shielding material. As a result of measuring the hydrogen content in the resin composition, the hydrogen content was about 9.8 wt % which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was 99.5 wt % or more, and the temperature at a residual weight ratio of 90 wt % was 370° C. or more, meaning that the composition exhibited extremely good heat resistance and heat stability. 1 g of a cationic polymerization initiator SI-80 (structural formula (11)) was added to a mixture of 8.23 g of a hydrogenated bisphenol A epoxy resin (YL6663, structural formula (14)), 8.85 g of a bisphenol A epoxy resin (Epicoat 828, structural formula (15)) and 8.85 g of an alicyclic epoxy resin (Celloxide 2021P, structural formula (8)) as epoxy resins. The mixture was sufficiently stirred until the polymerization initiator was dissolved, and then mixed with 50 g of copper as a density increasing agent to prepare a resin composition used for a neutron shielding material. As a result of measuring the hydrogen content in the resin composition, the hydrogen content was about 9.8 wt % which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was 99.5 wt % or more, and the temperature at a residual weight ratio of 90 wt % was 380° C. or more, meaning that the composition exhibited extremely good heat resistance and heat stability. 80.9 g of a hydrogenated bisphenol A epoxy resin (YL6663, structural formula (14)), 9.55 g of a bisphenol A epoxy resin (Epicoat 828, structural formula (15)) and 9.55 g of a polyfunctional alicyclic epoxy resin (EHPE3150, structural formula (7)) were mixed as epoxy resins. The mixture was maintained at 110° C. and sufficiently stirred until EHPE3150 (solid) was dissolved. After dissolution of EHPE3150, the mixture was allowed to stand in an environment at room temperature. When the temperature of the mixture was lowered to about room temperature, 1 g of a cationic polymerization initiator SI-80 (structural formula (11)) was added, and the mixture was sufficiently stirred until the polymerization initiator was dissolved. 50 g of copper was mixed therewith as a density increasing agent to prepare a resin composition used for a neutron shielding material. As a result of measuring the hydrogen content in the resin composition, the hydrogen content was about 9.8 wt % which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was 99.5 wt % or more, and the temperature at a residual weight ratio of 90 wt % was 390° C. or more, meaning that the composition exhibited extremely good heat resistance and heat stability. 77.3 g of a hydrogenated bisphenol A epoxy resin (YL6663, structural formula (14)), 11.35 g of an alicyclic epoxy resin (Celloxide 2021P, structural formula (8)) and 11.35 g of a polyfunctional alicyclic epoxy resin (EHPE3150, structural formula (7)) were mixed as epoxy resins. The mixture was maintained at 110° C. and sufficiently stirred until EHPE3150 (solid) was dissolved. After dissolution of EHPE3150, the mixture was allowed to stand in an environment at room temperature. When the temperature of the mixture was lowered to about room temperature, 1 g of a cationic polymerization initiator SI-80 (structural formula (11)) was added, and the mixture was sufficiently stirred until the polymerization initiator was dissolved. 50 g of copper was mixed therewith as a density increasing agent to prepare a resin composition used for a neutron shielding material. As a result of measuring the hydrogen content in the resin composition, the hydrogen content was about 9.8 wt % which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat. As a result, the residual weight ratio at 200° C. was 99.5 wt % or more, and the temperature at a residual weight ratio of 90 wt % was 390° C. or more, meaning that the composition exhibited extremely good heat resistance and heat stability. 80.38 g of a hydrogenated bisphenol A epoxy resin (YL6663, structural formula (14)), 6.54 g of a bisphenol A epoxy resin (Epicoat 828, structural formula (15)), 6.54 g of an alicyclic epoxy resin (Celloxide 2021P, structural formula (8)) and 6.54 g of a polyfunctional alicyclic epoxy resin (EHPE3150, structural formula (7)) were mixed as epoxy resins. The mixture was maintained at 110° C. and sufficiently stirred until EHPE3150 (solid) was dissolved. After dissolution of EHPE3150, the mixture was allowed to stand in an environment at room temperature. When the temperature of the mixture was lowered to about room temperature, 1 g of a cationic polymerization initiator SI-80 (structural formula (11)) was added, and the mixture was sufficiently stirred until the polymerization initiator was dissolved. 50 g of copper was mixed therewith as a density increasing agent to prepare a resin composition used for a neutron shielding material. As a result of measuring the hydrogen content in the resin composition, the hydrogen content was about 9.8 wt % which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was 99.5 wt % or more, and the temperature at a residual weight ratio of 90 wt % was 400° C. or more, meaning that the composition exhibited extremely good heat resistance and heat stability. 63.8 g of a hydrogenated bisphenol A epoxy resin (YL6663, structural formula (14)), 26.2 g of an alicyclic epoxy resin (Celloxide 2021P, structural formula (8)) as epoxy resins were mixed with 10 g of a hydrogenated bisphenol (manufactured by New Japan Chemical Co., Ltd., Rikabinol HB, structural formula (9)). The mixture was maintained at 100° C. and sufficiently stirred until Rikabinol HB (solid) was dissolved. After dissolution of Rikabinol HB, the mixture was allowed to stand in an environment at room temperature. When the temperature of the mixture was lowered to about room temperature, 1 g of a cationic polymerization initiator SI-80 (structural formula (11)) was added, and the mixture was sufficiently stirred until the polymerization initiator was dissolved. 50 g of copper was mixed therewith as a density increasing agent to prepare a resin composition used for a neutron shielding material. As a result of measuring the hydrogen content in the resin composition, the hydrogen content was about 9.8 wt % which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was about 99.5 wt %, and the temperature at a residual weight ratio of 90 wt % was 380° C. or more, meaning that the composition exhibited extremely good heat resistance and heat stability. 66.1 g of a hydrogenated bisphenol A epoxy resin (YL6663, structural formula (14)) and 23.9 g of an alicyclic epoxy resin (Celloxide 2021P, structural formula (8)) as epoxy resins were mixed with 10 g of cyclohexanedimethanol (manufactured by Tokyo Chemical Industry Co., Ltd., structural formula (10)). The mixture was maintained at 100° C. and sufficiently stirred until cyclohexanedimethanol (wax) was dissolved. After dissolution of cyclohexanedimethanol, the mixture was allowed to stand in an environment at room temperature. When the temperature of the mixture was lowered to about room temperature, 1 g of a cationic polymerization initiator SI-80 (structural formula (11)) was added, and the mixture was sufficiently stirred until the polymerization initiator was dissolved. 50 g of copper was mixed therewith as a density increasing agent to prepare a resin composition used for a neutron shielding material. As a result of measuring the hydrogen content in the resin composition, the hydrogen content was about 9.8 wt % which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat. As a result, the residual weight ratio at 200° C. was about 99.5 wt %, and the temperature at a residual weight ratio of 90 wt % was 380° C. or more, meaning that the composition exhibited extremely good heat resistance and heat stability. Here, evaluation was carried out for a neutron shielding material prepared by further mixing a neutron absorbent and a refractory material. 80.38 g of a hydrogenated bisphenol A epoxy resin (YL6663, structural formula (14)), 6.54 g of a bisphenol A epoxy resin (Epicoat 828, structural formula (15)), 6.54 g of an alicyclic epoxy resin (Celloxide 2021P, structural formula (8)) and 6.54 g of a polyfunctional alicyclic epoxy resin (EHPE3150, structural formula (7)) were mixed as epoxy resins. The mixture was maintained at 110° C. and sufficiently stirred until EHPE3150 (solid) was dissolved. After dissolution of EHPE3150, 39.0 g of copper as a density increasing agent, 76.0 g of magnesium hydroxide and 3.0 g of boron carbide were mixed therewith, and the mixture was stirred and maintained at 170° C. for 2 hours. After maintaining at 170° C. for 2 hours, the mixture was allowed to stand in an environment at room temperature. When the temperature of the mixture was about room temperature, 2 g of a cationic polymerization initiator SI-80 (structural formula (11)) was added, and the mixture was sufficiently stirred to prepare a neutron shielding material composition. The reference hydrogen content required for a neutron shielding material is a hydrogen content density of 0.096 g/cm3 or more. The hydrogen content density of the prepared neutron shielding material composition was measured to be 0.096 g/cm3 or more, which satisfied the standard. The hydrogen content in the resin component was separately measured to be 9.8 wt % or more. On the other hand, the resin composition for a neutron shielding material was cured at 170° C. for 4 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was 99.5 wt % or more, and the temperature at a residual weight ratio of 90 wt % was 400° C. or more, meaning that the composition exhibited extremely good heat resistance and heat stability. The cured product was enclosed in a closed vessel, and a thermal endurance test was carried out at 190° C. for 1,000 hours. The compressive strength was 1.4 times or more of that before the test, and the weight reduction was about 0.1%, meaning that the composition exhibited extremely good durability. 63.8 g of a hydrogenated bisphenol A epoxy resin (YL6663, structural formula (14)), 26.2 g of an alicyclic epoxy resin (Celloxide 2021P, structural formula (8)) as epoxy resins were mixed with 10 g of a hydrogenated bisphenol (Rikabinol HB, structural formula (9)). The mixture was maintained at 100° C. and sufficiently stirred until Rikabinol HB (solid) was dissolved. After dissolution of Rikabinol HB, 39.0 g of copper as a density increasing agent, 76.0 g of magnesium hydroxide and 3.0 g of boron carbide were mixed therewith, and the mixture was stirred and maintained at 170° C. for 2 hours. After maintaining at 170° C. for 2 hours, the mixture was allowed to stand in an environment at room temperature. When the temperature of the mixture was about room temperature, 2 g of a cationic polymerization initiator SI-80L (structural formula (11)) was added, and the mixture was sufficiently stirred to prepare a neutron shielding material composition. The reference hydrogen content required for a neutron shielding material is a hydrogen content density of 0.096 g/cm3 or more. The hydrogen content density of the prepared neutron shielding material composition was measured to be 0.096 g/cm3 or more, which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 170° C. for 4 hours to measure the weight reduction by heat. As a result, the residual weight ratio at 200° C. was about 99.5 wt %, and the temperature at a residual weight ratio of 90 wt % was 380° C. or more, meaning that the composition exhibited extremely good heat resistance and heat stability. The cured product was enclosed in a closed vessel, and a thermal endurance test was carried out at 200° C. for 500 hours. The compressive strength was 1.2 times or more of that before the test, and the weight reduction was about 0.1%, meaning that the composition exhibited extremely good durability. Next, performance of neutron shielding materials employing a conventionally used composition not containing a density increasing agent was evaluated. A refractory material or neutron absorbent was not added as in Examples. The hydrogen content was determined by component analysis, and the weight reduction by heat was determined by measurement using TGA. 82.5 g of a hydrogenated bisphenol A epoxy resin as in Example 1 represented by the structural formula (14) (Yuka Shell Epoxy K.K., YL6663) as an epoxy resin and 17.5 g of isophoronediamine as a curing agent were sufficiently stirred to prepare a resin composition used for a neutron shielding material. This is a comparative example in which the present invention is compared with a neutron absorbent employed a curing agent. A density increasing agent was not added. As a result of measuring the hydrogen content in the resin composition, the hydrogen content was 9.8 wt % or more which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was about 99.5 wt %, and the temperature at a residual weight ratio of 90 wt % was about 300° C., meaning that the composition exhibited heat resistance and heat stability inferior to those of the compositions of Examples. This composition system considerably differs from that in Example 1 in that an amine curing agent is used instead of a cationic polymerization initiator. As is clear from comparison of the composition of Example 1 with the composition of Comparative Example 1, heat resistance and heat stability are improved by curing with a polymerization initiator as in Example 1. 81.4 g of a bisphenol A epoxy resin (Epicoat 828, structural formula (15)) as an epoxy resin and 18.6 g of isophoronediamine as a curing agent were sufficiently stirred to prepare a resin composition used for a neutron shielding material. A density increasing agent was not added. As a result of measuring the hydrogen content in the resin composition, the hydrogen content was 8.2 wt % or less which was considerably below the standard, unsatisfactorily. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was about 99.5 wt %, and the temperature at a residual weight ratio of 90 wt % was about 350° C., meaning that the composition exhibited good heat resistance and heat stability. This composition system has good heat resistance and heat stability, but is not suitable as a resin composition for a neutron shielding material in terms of hydrogen content. This composition system considerably differs from that in Example 2 in that an amine curing agent is used instead of a cationic polymerization initiator. As is also clear from comparison of the composition of Comparative Example 2 with the composition of Comparative Example 3, heat resistance and heat stability are improved by curing with a polymerization initiator. A bisphenol A epoxy resin (Epicoat 828, structural formula (15)) as an epoxy resin was mixed with a polyamine curing agent at a mixing ratio of 1:1 (stoichiometrically equal), and the mixture was stirred to prepare a resin composition used for a neutron shielding material. A density increasing agent was not added. As a result of measuring the hydrogen content in the resin composition, the hydrogen content was 9.8 wt % or more which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was about 99 wt % or less, and the temperature at a residual weight ratio of 90 wt % was 300° C. or less, meaning that the composition exhibited heat resistance and heat stability inferior to those of the compositions of Examples. This composition system imitates the same system as in a conventionally used resin composition for a neutron shielding material. The composition of Comparative Example 4 is suitable in terms of hydrogen content, but has low heat resistance and heat stability as compared with those of the compositions of Examples. It can be found that the compositions of Examples have excellent heat resistance and heat stability. 81.7 g of an epoxy resin having a structure in which OH at each end of polypropylene glycol is substituted with glycidyl ether (epoxy equivalent: 190) and 18.3 g of isophoronediamine as a curing agent were sufficiently stirred to prepare a resin composition used for a neutron shielding material. A density increasing agent was not added. As a result of measuring the hydrogen content in the resin composition, the hydrogen content was 9.8 wt % or more which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was 99.5 wt % or less, and the temperature at a residual weight ratio of 90 wt % was 250° C. or less, meaning that the composition exhibited heat resistance and heat stability extremely inferior to those of the compositions of Examples. 78.5 g of 1,6-hexane diglycidyl ether (epoxy equivalent: 155) as an epoxy resin and 21.5 g of isophoronediamine as a curing agent were sufficiently stirred to prepare a resin composition used for a neutron shielding material. A density increasing agent was not added. As a result of measuring the hydrogen content in the resin composition, the hydrogen content was 9.8 wt % or more which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was 99.5 wt % or less, and the temperature at a residual weight ratio of 90 wt % was 300° C. or less, meaning that the composition exhibited heat resistance and heat stability inferior to those of the compositions of Examples. Here, a neutron absorbent was added to a conventional resin component to evaluate the neutron shielding effect. 50 g of a bisphenol A epoxy resin (Epicoat 828, structural formula (15)) as an epoxy resin was mixed with 50 g of a polyamine curing agent, and the mixture was stirred. 146.5 g of magnesium hydroxide and 3.5 g of boron carbide were mixed therewith, and the mixture was stirred to prepare a resin composition for a neutron shielding material. A density increasing agent was not added. The reference hydrogen content required for a neutron shielding material is a hydrogen content density of 0.096 g/cm3 or more. The hydrogen content density of the prepared neutron shielding material composition was measured to be 0.096 g/cm3 or more, which satisfied the standard. On the other hand, the resin composition for a neutron shielding material was cured at 80° C. for 30 minutes and at 150° C. for 2 hours to measure the weight reduction by heat in the same manner as in Example 1. As a result, the residual weight ratio at 200° C. was about 99 wt % or less, and the temperature at a residual weight ratio of 90 wt % was 300° C. or less, meaning that the composition exhibited heat resistance and heat stability inferior to those of the compositions of Examples. The cured product was enclosed in a closed vessel, and a thermal endurance test was carried out at 190° C. for 1,000 hours. The compressive strength was decreased by 30% or more as compared with that before the test, meaning that the composition has low durability in a high-temperature environment. This composition system imitates the same system as in a conventionally used neutron shielding material composition. The composition of Comparative Example 6 is suitable in terms of hydrogen content, but has low heat resistance and heat stability as compared with those of the compositions of Examples 11 and 12. It can be found that the compositions of Examples have excellent heat resistance and heat stability. As is clear from the above Examples and Comparative Examples, resins cured with the polymerization initiator of the present invention have a temperature at a residual weight ratio of 90 wt % increased by 30 to 50° C. on average as compared with resins using the same polymerization component cured with an amine curing agent, and such resins has high heat resistance. A neutron shielding material is obtained from the neutron shielding material composition of the present invention by curing a heat-resistant polymerization component with a cationic polymerization initiator. When a shielding material is prepared by curing the composition of the present invention polymerizable without using a curing agent component that has a bond easily decomposed under high-temperature conditions, the shielding material has an increased heat-resistant temperature and has ensured neutron shielding effect. Accordingly, the present invention can provide a composition for a neutron shielding material that can endure long-term storage of spent nuclear fuels. Further, since the composition of the present invention comprises a density increasing agent, the neutron shielding material can provide an increased neutron absorption while maintaining secondary γ-ray shielding performance.
	claims	1. A waste vitrification method performed at least in part in a vessel having a wall and a rotatable impeller comprising the steps of:introducing a feed stream comprising waste material into the vessel; mixing the feed stream into a glass melt formed in the vessel with the impeller to disperse said feed stream in the melt to form a foamy mass, said foamy mass comprising gaseous material released by the waste material into the glass melt; completing an electrical circuit between the wall of the vessel and the impeller and including said foamy mass to form a molten vitrified output; densifying the foamy mass by passing said foamy mass into a quiescent zone where a portion of the gaseous material in said foamy mass separates from said foamy mass to form a molten vitrifiable output; and recovering the molten vitrified output. 2. The waste vitrification method of claim 1 wherein said feed stream further comprises glass batch in solid form. 3. The waste vitrification method of claim 1 wherein said feed stream is an aqueous mixture. 4. The waste vitrification method of claim 3 wherein said feed stream is approximately 60 percent water and 40 percent solids by weight. 5. The waste vitrification method of claim 4 wherein said solids include frit. 6. The waste vitrification method of claim 2 wherein said feed stream comprises a first feed stream of waste materials and a second feed stream of glass batch in solid form. 7. The waste vitrification method of claim 1 wherein said step of mixing is performed by an impeller having outer surface portion moving at a speed of more than 250 feet per minute. 8. The waste vitrification method of claim 1 wherein said step of mixing is performed by an impeller having outer surface portion moving at a speed of more than 500 feet per minute. 9. The waste vitrification method of claim 1 wherein said steps of mixing and heating the foamed material are performed simultaneously in the vessel. 10. The method of claim 1 wherein said heating continues through the densifying step to maintain the molten vitrified output in a pourable state until after the recovering step. 11. The the method of claim 1, further comprising the step of cooling the molten vitrified output to form a solidified vitreous mass. 12. The method of claim 1 wherein said feed stream includes radioactive material. 13. The method of claim 1 wherein said glass batch material is boro-silicate glass. 14. A waste vitrification method performed at least in part in a vessel having a wall and a rotatable impeller comprising the steps of:introducing a mixture of solid waste material and glass batch into the vessel; mixing the mixture in the vessel with the impeller to form a foamy mass comprising gaseous material released by the waste material into the glass melt; completing an electrical circuit from the wall of the vessel through said foamy mass to said impeller to provide heat which forms a molten vitrified output; densifying the foamy mass by passing the foamy mass into a quiescent zone where the foamy mass separates into a densified portion and a gaseous portion; and recovering the densified portion. 15. The waste vitrification method of claim 14 wherein said steps of mixing and heating the foam material are performed simultaneously in the vessel. 16. The method of claim 14 wherein residual heat is maintained through the densifying step to keep the densified material in a pourable state until after the recovering step. 17. A waste vitrification method performed at least in part in the vessel having a wall and a rotatable impeller comprising the steps of:introducing a mixture of solid waste material and glass batch material into the vessel; mixing the mixture in the vessel with the impeller to form a foamy mass comprising solid waste material, glass batch material and gaseous material released by said solid waste material and glass batch material; completing an electrical series circuit relationship including the wall of the vessel, the foamy mass and the impeller; densifying the foamy mass by passing the foamy mass into a quiescent zone where the foamy mass separates into a densified portion and a gaseous portion; and recovering the densified portion.
050283804	claims	1. A method for the identification of the leakiness of a neutron-absorbing pencil of a nuclear reactor, said pencil having a hollow tube and a core formed by at least one metal, said method comprising the steps of: placing said pencil in an impervious chamber filled with a chemical solution capable of reacting with said metal; putting said solution under pressure in order to make the solution penetrate the pencil through the leakiness fault of the presumably defective pencil; then lowering said pressure to enable the solution to go out of the presumably defective pencil in the impervious chamber; performing an analysis of said solution in order to enable the showing up, in the solution, of chemical derivatives of said metals of the core of said pencil. a step consisting in the measurement of the chemical concentration of silver, indium, cadmium in the solution before the pencil is placed in the chamber; the repeating of this measuring step after the pressurizing operation. measuring the chemical concentration of silver, indium, cadmium in the solution before the pencil is placed in the chamber; and repeating the measuring step after the heating process. 2. A method according to claim 1, wherein the aggressive chemical solution is a solution of an acid taken from the following group: nitric acid, sulphuric acid, hydrochloric acid. 3. A method according to claim 1 wherein, for a pencil with its core formed by an alloy of cadmium, indium and silver, said chemical derivatives are the metallic salts of the constituent elements of the alloy. 4. A method according to claim 1 wherein, in addition to the showing up of said chemical derivatives, the method consists in performing, at the solution, after lowering the pressure, the detection of the radioactive isotopes of an alloy, notably silver Ag 110 m, with reference to a threshold value of concentration. 5. A method according to claim 4, wherein the detection of the radioactive isotopes of the alloy, notably of silver Ag 110 m is done by gamma spectrometry. 6. A method according to claim 5 wherein, prior to the step in which said pencil is placed in the impervious chamber filled with said chemical solution, which forms the starting solution, said method consists in the determining, by spectrometry, of the concentrations in radioactive isotopes, including Ag 110 m, of the starting solution, said concentrations of the starting solution being chosen as a threshold value of concentration. 7. A method according to claim 1, wherein said pressure applied to the solution in order to make the solution penetrate the pencil is maintained for a period of at least ten minutes. 8. A method according to claim 7, further comprising: 9. A method according to claim 1, wherein said solution, before or after the insertion of said pencil in it, is subjected to a heating process used to accelerate a chemical reaction of the solution on the constituent elements of the core of said pencil. 10. A method according to claim 9, further comprising the steps of:
	summary
042591546	abstract	A nuclear reactor containment structure including a diaphragm floor for dividing a closure casing into a drywell and a pressure suppression chamber, the diaphragm floor being supported at its inner peripheral end by a pedestal for a pressure vessel and abutting at its outer peripheral end against shear keys secured to the closure casing. The diaphragm floor includes a concrete layer embedded between structural steels extending radially between the pedestal and the shear keys.
044735294	summary	FIELD OF THE INVENTION The invention relates to a device for collecting purge liquids and gases in an installation containing substances which may possess a certain degree of radioactivity, such as the shut-down cooling circuit of a pressurized water nuclear reactor. BACKGROUND In a nuclear power station, it is necessary to be able to purge the circuits which are not used in a continuous manner, such as the shut-down cooling circuit and, more generally, all the parts of the circuits which can be isolated. It is also necessary to be able to purge those enclosures which are likely to hold liquids or gases which are contaminated by radioactive substances coming from the reactor core. This purging operation can be carried out during cold close-down periods, and requires certain precautions to be taken, as the liquids or gases remaining in the circuits being purged are generally contaminated by radioactive substances coming from the reactor core. At present, collecting networks are used which are connected directly to the various components of the piping system and to the pieces of equipment which are being purged. These networks are provided with inspection windows which permit visual inspection of the flow during purging and are connected in a permanent manner to the circuits from which they are intended to collect the purged liquids and gases. As the points in the piping system where purging or drainage has to be carried out are many in number and are distributed throughout the circuit to be purged, the circuit for collecting the removed liquids and gases is generally complicated and of appreciable length. The length of piping constituting this network for collecting such liquids and gases rapidly becomes contaminated since it is only rarely that fluids pass through them, and thus conditions are created under which it is fairly easy for radioactive deposits to be formed there. Some of these pipes become contaminated to such a degree that they have to be replaced during the life of the power station. When the drainage of the liquids contained in the piping or the equipment to be purged has been completed, the section of piping which has been purged is filled with air, which, when the piping system is being filled, escapes through vents, carrying with it a small amount of liquid which contains radioactive products. It is consequently necessary to send this air through a phase separator before releasing the gaseous phase into the atmosphere of the building. The collected contaminated liquids are sent to an installation for treating the contaminated effluent in which the radioactive products are eliminated. Installations for collecting the drained liquids and gases in the circuits constituting the equipment of nuclear power stations are consequently very complex and involved and increase the constructional costs of the latter. The object of the invention is consequently to provide a device for collecting the purged liquid and gases in an installation containing substances which may possess a certain degree of radioactivity, and including closable purging means distributed at different locations in the installation, such device making it possible to avoid the use of complex piping networks connected in a permanent manner to the installation. In order to achieve this object, the self-contained and movable collecting device comprises: (a) a removable and transportable enclosure of small dimensions which can be connected to the purging means of the installation, in such a way that the internal volume of the enclosure is isolated from the external medium; (b) a means for opening or closing the purging means which can be operated from outside the enclosure for isolating the inside of the enclosure from the installation to be purged or putting these two in communication; (c) a movable unit for drawing off and collecting the purged liquids and gases, including at least one collecting reservoir, a suction means arranged inside this reservoir, a means for evacuating gases into the atmosphere and a connecting component for evacuation of the purged liquids into a treatment installation; (d) a flexible connecting pipe between the enclosure and the drawing-off and collecting unit. A description will now be given, with reference to the attached drawings, of one embodiment of a device for collecting purge liquid and gases which can for example, be used in a pressurized water nuclear power station, for purging the cooling circuit of the shut-down reactor.
	summary
	claims	1. A method for planning delivery of radiation dose to a target area within a subject, the method comprising:defining a set of one or more optimization goals, the set of one or more optimization goals comprising a desired dose distribution in the subject;specifying an initial plurality of control points along an initial trajectory, the initial trajectory involving relative movement between a radiation source and the subject in a source trajectory direction; anditeratively optimizing a simulated dose distribution relative to the set of one or more optimization goals to determine one or more radiation delivery parameters associated with each of the initial plurality of control points;wherein, for each of the initial plurality of control points, the one or more radiation delivery parameters comprise positions of a plurality of leaves of a multi-leaf collimator (MLC), the plurality of leaves moveable in a leaf-translation direction; andwherein during relative movement between the radiation source and the subject along the initial trajectory, the leaf-translation direction is oriented at a MLC orientation angle φ with respect to the source trajectory direction and wherein an absolute value of the MLC orientation angle φ satisfies 0°<\|φ\|<90°. 2. A planning method according to claim 1 wherein the absolute value of the MLC orientation angle φ satisfies 15°<=\|φ\|<=75°. 3. A planning method according to claim 1 wherein the absolute value of the MLC orientation angle φ satisfies 30°<=\|φ\|<=60°. 4. A planning method according to claim 1 wherein the MLC orientation angle φ is constant throughout the initial trajectory. 5. A planning method according to claim 3 wherein the MLC orientation angle φ is constant throughout the initial trajectory. 6. A planning method according to claim 3 wherein the initial trajectory comprises at least one pair of locations wherein a first beam directed from the radiation source toward the subject from a first one of the pair of locations and a second beam directed from the radiation source toward the subject from a second one of the pair of locations are substantially parallel but opposing one another. 7. A planning method according to claim 3 wherein the initial trajectory comprises a plurality of arcs, each arc involving relative movement between the radiation source and the subject within a corresponding plane. 8. A planning method according to claim 7 wherein, between successive ones of the plurality of arcs, the initial trajectory comprises inter-arc relative movement between the radiation source and the subject, the inter-arc relative movement comprising movement such that the corresponding planes associated with each arc intersect one another. 9. A planning method according to claim 7 wherein, between successive ones of the plurality of arcs, the initial trajectory comprises inter-arc relative movement between the radiation source and the subject, the inter-arc relative movement comprising movement such that the corresponding planes associated with each arc are parallel with one another. 10. A planning method according to claim 3 comprising, upon reaching one or more initial termination conditions:adding one or more additional control points to obtain an increased plurality of control points;iteratively optimizing the simulated dose distribution relative to the set of optimization goals to determine one or more radiation delivery parameters associated with each of the increased plurality of control points. 11. A planning method according to claim 1 wherein the initial trajectory comprises a plurality of arcs, each arc involving relative movement between the radiation source and the subject within a corresponding plane. 12. A planning method according to claim 11 wherein, between successive ones of the plurality of arcs, the initial trajectory comprises inter-arc relative movement between the radiation source and the subject, the inter-arc relative movement comprising movement such that the corresponding planes associated with each arc intersect one another. 13. A planning method according to claim 11 wherein, between successive ones of the plurality of arcs, the initial trajectory comprises inter-arc relative movement between the radiation source and the subject, the inter-arc relative movement comprising movement such that the corresponding planes associated with each arc are parallel with one another. 14. A planning method according to claim 1 comprising, upon reaching one or more initial termination conditions:adding one or more additional control points to obtain an increased plurality of control points;iteratively optimizing the simulated dose distribution relative to the set of optimization goals to determine one or more radiation delivery parameters associated with each of the increased plurality of control points. 15. A planning method according to claim 1 wherein the initial trajectory comprises a first non-self overlapping trajectory which involves non-self overlapping relative movement between the radiation source and the subject, wherein the initial plurality of control points comprises a first plurality of control points along the first non-self overlapping trajectory and wherein iteratively optimizing the simulated dose distribution relative to the set of one or more optimization goals comprises iteratively optimizing the simulated dose distribution relative to the set of one or more optimization goals over the first plurality of control points along the first non-self overlapping trajectory to determine one or more radiation delivery parameters associated with each of the first plurality of control points along the first non-self overlapping trajectory. 16. A planning method according to claim 15 wherein the initial trajectory comprises a second non-self overlapping trajectory which involves non-self overlapping relative movement between the radiation source and the subject, wherein the initial plurality of control points comprises a second plurality of control points along the second non-self overlapping trajectory and wherein iteratively optimizing the simulated dose distribution relative to the set of one or more optimization goals comprises iteratively optimizing the simulated dose distribution relative to the set of one or more optimization goals over the second plurality of control points along the second non-self overlapping trajectory to determine one or more radiation delivery parameters associated with each of the second plurality of control points along the second non-self overlapping trajectory. 17. A planning method according to claim 16 wherein the first and second non-self overlapping trajectories overlap one another over at least a portion thereof. 18. A method for planning delivery of radiation dose to a target area within a subject, the method comprising:defining a set of one or more optimization goals, the set of one or more optimization goals comprising a desired dose distribution in the subject;specifying an initial plurality of control points along an initial trajectory, the initial trajectory involving relative movement between a radiation source and the subject in a source trajectory direction; anditeratively optimizing a simulated dose distribution relative to the set of one or more optimization goals to determine one or more radiation delivery parameters associated with each of the initial plurality of control points;wherein, for each of the initial plurality of control points, the one or more radiation delivery parameters comprise positions of a plurality of leaves of a multi-leaf collimator (MLC), the plurality of leaves moveable in a leaf-translation direction; andwherein during relative movement between the radiation source and the subject along the initial trajectory, the leaf-translation direction is oriented at a MLC orientation angle φ with respect to the source trajectory direction and wherein an absolute value of the MLC orientation angle φ satisfies 0°<\|φ\|<90°;wherein the initial trajectory comprises at least one pair of locations wherein a first beam directed from the radiation source toward the subject from a first one of the pair of locations and a second beam directed from the radiation source toward the subject from a second one of the pair of locations are substantially parallel but opposing one another. 19. A method for delivering radiation dose to a target area within a subject, the method comprising:defining a trajectory for relative movement between a treatment radiation source and the subject in a source trajectory direction;determining a radiation delivery plan;while effecting relative movement between the treatment radiation source and the subject along the trajectory in the source trajectory direction, delivering a treatment radiation beam from the treatment radiation source to the subject according to the radiation delivery plan to impart a dose distribution on the subject;wherein delivering the treatment radiation beam from the treatment radiation source to the subject comprises varying an intensity of the treatment radiation beam over at least a portion of the trajectory. 20. A radiation delivery method according to claim 19 wherein the trajectory comprises a plurality of arcs, each arc involving relative movement between the radiation source and the subject within a corresponding plane. 21. A radiation delivery method according to claim 20 wherein, between successive ones of the plurality of arcs, the trajectory comprises inter-arc relative movement between the radiation source and the subject, the inter-arc relative movement comprising movement such that the corresponding planes associated with each arc intersect one another. 22. A radiation delivery method according to claim 20 wherein, between successive ones of the plurality of arcs, the trajectory comprises inter-arc relative movement between the radiation source and the subject, the inter-arc relative movement comprising movement such that the corresponding planes associated with each arc are parallel with one another. 23. A radiation delivery method according to claim 19 wherein varying the intensity of the treatment radiation beam over at least the portion of the trajectory comprises varying a rate of radiation output of the radiation source while effecting continuous relative movement between the treatment radiation source and the subject along the trajectory. 24. A radiation delivery method according to claim 23 wherein the trajectory comprises a plurality of arcs, each arc involving relative movement between the radiation source and the subject within a corresponding plane. 25. A radiation delivery method according to claim 24 wherein, between successive ones of the plurality of arcs, the trajectory comprises inter-arc relative movement between the radiation source and the subject, the inter-arc relative movement comprising movement such that the corresponding planes associated with each arc are parallel with one another. 26. A radiation delivery method according to claim 24 wherein, between successive ones of the plurality of arcs, the trajectory comprises inter-arc relative movement between the radiation source and the subject, the inter-arc relative movement comprising movement such that the corresponding planes associated with each arc intersect one another. 27. A radiation delivery method according to claim 19 wherein delivering the treatment radiation beam from the treatment radiation source to the subject comprises varying a shape of the treatment radiation beam over at least the portion of the trajectory. 28. A radiation delivery method according to claim 27 wherein varying the shape of the treatment radiation beam over at least the portion of the trajectory, comprises varying positions of a plurality of leaves of a multi-leaf collimator (MLC), the plurality of leaves moveable in a leaf-translation direction and wherein during relative movement between the treatment radiation source and the subject along the trajectory, the leaf-translation direction is oriented at a MLC orientation angle φ with respect to the source trajectory direction and wherein an absolute value of the MLC orientation angle φ satisfies 0°<\|φ\|<90°. 29. A radiation delivery method according to claim 28 wherein the absolute value of the MLC orientation angle φ satisfies 15°<=\|φ\|<=75°. 30. A radiation delivery method according to claim 28 wherein the absolute value of the MLC orientation angle φ satisfies 30°<=\|φ\|<=60°. 31. A radiation delivery method according to claim 30 wherein the MLC orientation angle φ is constant throughout the initial trajectory. 32. A radiation delivery method according to claim 30 wherein the trajectory comprises at least one pair of locations wherein a first treatment radiation beam directed from the radiation source toward the subject from a first one of the pair of locations and a second treatment radiation beam directed from the radiation source toward the subject from a second one of the pair of locations are substantially parallel but opposing one another. 33. A radiation delivery method according to claim 28 wherein the MLC orientation angle φ is constant throughout the initial trajectory. 34. A radiation delivery method according to claim 28 wherein the trajectory comprises a plurality of arcs, each arc involving relative movement between the radiation source and the subject within a corresponding plane. 35. A radiation delivery method according to claim 34 wherein, between successive ones of the plurality of arcs, the trajectory comprises inter-arc relative movement between the radiation source and the subject, the inter-arc relative movement comprising movement such that the corresponding planes associated with each arc intersect one another. 36. A radiation delivery method according to claim 34 wherein, between successive ones of the plurality of arcs, the trajectory comprises inter-arc relative movement between the radiation source and the subject, the inter-arc relative movement comprising movement such that the corresponding planes associated with each arc are parallel with one another. 37. A method for delivering radiation dose to a target area within a subject, the method comprising:defining a trajectory for relative movement between a treatment radiation source and the subject in a source trajectory direction;determining a radiation delivery plan;while effecting relative movement between the treatment radiation source and the subject along the trajectory in the source trajectory direction, delivering a treatment radiation beam from the treatment radiation source to the subject according to the radiation delivery plan to impart a dose distribution on the subject;wherein delivering the treatment radiation beam from the treatment radiation source to the subject comprises varying at least one of: an intensity of the treatment radiation beam; and a shape of the treatment radiation beam over at least a portion of the trajectory;wherein varying at least one of the intensity of the treatment radiation beam and the shape of the treatment radiation beam over at least the portion of the trajectory, comprises varying positions of a plurality of leaves of a multi-leaf collimator (MLC), the plurality of leaves moveable in a leaf-translation direction and wherein during relative movement between the treatment radiation source and the subject along the trajectory, the leaf-translation direction is oriented at a MLC orientation angle φ with respect to the source trajectory direction and wherein an absolute value of the MLC orientation angle φ satisfies 0°<\|φ\|<90°; andwherein the trajectory comprises at least one pair of locations wherein a first treatment radiation beam directed from the radiation source toward the subject from a first one of the pair of locations and a second treatment radiation beam directed from the radiation source toward the subject from a second one of the pair of locations are substantially parallel but opposing one another. 38. A method for delivering radiation dose to a target area within a subject, the method comprising:defining a trajectory for relative movement between a treatment radiation source and the subject in a source trajectory direction;determining a radiation delivery plan;while effecting relative movement between the treatment radiation source and the subject along the trajectory in the source trajectory direction, delivering a treatment radiation beam from the treatment radiation source to the subject according to the radiation delivery plan to impart a dose distribution on the subject;wherein delivering the treatment radiation beam from the treatment radiation source to the subject comprises varying at least one of: an intensity of the treatment radiation beam; and a shape of the treatment radiation beam over at least a portion of the trajectory;wherein varying at least one of the intensity of the treatment radiation beam and the shape of the treatment radiation beam over at least the portion of the trajectory comprises varying a rate of radiation output of the radiation source while effecting continuous relative movement between the treatment radiation source and the subject along the trajectory; andwherein the trajectory comprises at least one pair of locations wherein a first treatment radiation beam directed from the radiation source toward the subject from a first one of the pair of locations and a second treatment radiation beam directed from the radiation source toward the subject from a second one of the pair of locations are substantially parallel but opposing one another.
	abstract	Various embodiments of a nuclear radiation particle power converter and method of forming such power converter are disclosed. In one or more embodiments, the power converter can include first and second electrodes, a three-dimensional current collector disposed between the first and second electrodes and electrically coupled to the first electrode, and a charge carrier separator disposed on at least a portion of a surface of the three-dimensional current collector. The power converter can also include a hole conductor layer disposed on at least a portion of the charge carrier separator and electrically coupled to the second electrode, and nuclear radiation-emitting material disposed such that at least one nuclear radiation particle emitted by the nuclear radiation-emitting material is incident upon the charge carrier separator.
048062782	abstract	The invention relates to a method of and apparatus for the regregation of radioactive iodine isotopes from fluid samples. The method consists in leading the degassed and in certain cases also otherwise prepared sample into a column filled with an amorphous solid medium for binding quantitatively the cations, fluoride anions and contaminations of colloidal state, then the effluent flow continuously through an energy-selective gamma detector and continuously recording the signals generated in the detector by the radioactive iodine isotopes. The proposed apparatus comprises a sampling unit (MV), a degassing vessel (G), fluid transfer pumps (P1, P2), pipings, cocks (CS), a bubble removing cell (BC), a column (K) with amorphous material filling, an energy-selective detector system (GD) for measuring gamma radiation, and a signal processing and recording unit (JR).
	abstract	An example electron beam drawing apparatus includes an electron beam emitting unit which emits an electron beam, a rotary stage which rotatably supports a turntable for retaining a drawing object, and a sample stage which is supported by the turntable in a range including a rotating center of the turntable to retain an adjustment sample. A rotationally symmetrical pattern such as a concentric pattern and a radial pattern can be drawn in the drawing object by irradiating the drawing object with the electron beam during rotation of the turntable. Before the pattern is actually drawn in the drawing object, beam adjustment and rotating center adjustment are performed using an adjustment sample. The adjustment sample is retained by the sample stage, and the sample stage is supported by the turntable in the range including the rotating center of the turntable. Therefore, the beam adjustment and the rotating center adjustment can be performed using the adjustment sample supported by the turntable, and the different stage for placing the adjustment sample is not required, which allows the apparatus to be miniaturized.
	abstract	A system that monitors telemetry from a host computer system to detect degradation in a remote storage device. During operation, the system monitors performance parameters from a host computer system which accesses the remote storage device, wherein the performance parameters relate to the interactions between the host computer system and the remote storage device. The system then determines whether the monitored performance parameters have deviated from predicted values for the performance parameters. If so, the system generates a signal indicating that the remote storage device has degraded.
	claims	1. A local diagnostic subsystem usable to diagnose a component of a physical system, comprising: at least one observation variable, each observation variable storing a physical value received from a sensor monitoring a first component that describes a state of the first component; at least one assumption variable, each assumption variable storing a value representing an assumption about the state of the first component; at least one dependent variable, each dependent variable storing a value representing a quantity relating one of the at least one observation variable to one of the at least one assumption variable; a plurality of constraints, each constraint relating two of the at least one observation variable, the at least one assumption variable and the at least one dependent variable to describe a behavior of the component; and at least one diagnosis variable relating a local diagnosis of the local diagnostic subsystem to a second local diagnosis of a second local diagnostic subsystem. 2. The local diagnostic subsystem of claim 1 , wherein the one or more variables comprise at least one of at least one local pseudo-assumption variable that corresponds to a pseudo-assumption variable of the second local diagnostic subsystem, and at least one local pseudo-observation variable that corresponds to a pseudo-assumption variable of the second local diagnostic subsystem. claim 1 3. A diagnostic system usable to diagnose a component of a physical system, comprising: an initialization circuit, routine or application that initializes the diagnosis system to an initial diagnosis; a local diagnosis circuit, routine or application that generates one or more candidate local diagnoses based on a physical value received from a sensor monitoring the component, the physical value describing a state of the component; a diagnosis management circuit, routine or application that selects a local diagnosis from among the one or more candidate local diagnoses; and a remote diagnosis management circuit, routine or application that communicates with a second diagnostic system, wherein zero, one or more of the candidate local diagnoses are based in part on a second local diagnosis generated by the second diagnostic system. 4. A method for diagnosing a physical system, the system comprising a plurality of physical components being diagnosed by a plurality of local diagnostic subsystems, the method comprising: initializing a first local diagnostic subsystem to an initial diagnosis; assigning an observation value to an observation variable of the first local diagnostic subsystem, wherein the observation value corresponds to a physical value received from a sensor monitoring a first component and describing a state of the first component; performing a first local diagnosis in the first local diagnostic subsystem to determine a minimal cost diagnosis explaining the observation value; determining whether the lowest cost diagnosis requires a change in the value of a pseudo-assumption variable of the first local diagnostic subsystem to a desired value; performing a second local diagnosis in the first local diagnostic subsystem in which the value of the pseudo-assumption variable is held constant to determine a maximum cost; supplying a desired value and the determined maximum cost to a second local diagnostic subsystem where the second local diagnostic subsystem contains a pseudo-observation variable corresponding to the pseudo-assumption variable of the first local diagnostic subsystem; performing a third local diagnosis in the second local diagnostic subsystem to determine whether a diagnosis in which the pseudo-observation variable is set to the desired value can be performed for a cost less than the maximum cost; adopting the first local diagnosis in the first local diagnostic subsystem when the third local diagnosis can not be performed for a cost less than the maximum cost; and adopting the second local diagnosis in the first local diagnostic subsystem when the third local diagnosis can be performed for a cost less than the maximum cost.
	claims	1. An electron beam irradiation device for irradiating an electron beam to a belt-shaped irradiated object while making the irradiated object travel,an electron beam generating section which generates the electron beam and emits the electron beam to an outside from a transmission window part;an irradiation chamber adjacent to the transmission window part of the electron beam generating section, having partitions surrounding a periphery, a feed-in opening which opens on the partition to allow the belt-shaped irradiated object to be fed in, and a feed-out opening which opens on the partition to allow the belt-shaped irradiated object to be fed out, and formed as a closed space filled with inert gas, in which the electron beam emitted from the transmission window section is irradiated to the belt-shaped irradiated object fed in from an outside and travels the inside; andan oxygen cutoff section adjacent to the irradiation chamber on an upstream side in an irradiated object traveling direction, having a feed-in opening for feeding in the belt-shaped irradiated object, and a feed-out opening for feeding out the belt-shaped irradiated object, and formed as a closed space, in which the belt-shaped irradiated object travels to be introduced to the irradiation chamber, the inert gas is blown to the irradiating surface side of the irradiated object, and oxygen in the air accompanying a vicinity of a surface of the irradiated object to flow in is shut off or diluted, wherein:the oxygen cutoff section surrounds the irradiated object with a surface side partition facing a side of the irradiating surface of the traveling belt-shaped irradiated object, a backface side partition facing to a side opposite to the irradiating surface of the irradiated object, and a pair of sideface side partitions facing both sideface sides of the irradiated object,a gap Ws between the surface side partition and the backface side partition of the oxygen cutoff section, and a gap We between the surface side partition and the backface side partition of the irradiation chamber and across the belt-shaped irradiated object in the irradiation chamber satisfy an inequality Ws<We,the gap Ws between the surface side partition and the backface side partition of the oxygen cutoff section is made uniform or almost uniform throughout an entire area of the oxygen cutoff section and,a blowing slit for the inert gas is provided on the surface side partition of the oxygen cutoff section, with a blowing opening thereof being not projected form or caved in the surface side partition of the oxygen cutoff section. 2. The electron beam irradiation device according to claim 1, further provided with a coating part for coating a liquid electron beam curing resin in a non-curing state on the surface of the irradiated object on the upstream side in the irradiated object traveling direction in the oxygen cutoff section. 3. The electron beam irradiation device according to claim 1, wherein the gap Ws between the surface side partition and the backface side partition of the oxygen cutoff section is set to be wider than a thickness of the irradiated object by a range of 1-20 mm. 4. The electron beam irradiation device according to claim 1, wherein the slit is formed so that a blowing direction of the inert gas from the slit inclines toward the upstream side in the traveling direction relative to a direction perpendicular to the traveling direction of the irradiated object. 5. The electron beam irradiation device according to claim 1, wherein on a downstream side relative to the slit in the traveling direction of the irradiated object, a gas supplying hole for supplying the inert gas for the irradiated object from the same side as the slit is provided. 6. The electron beam irradiation device according to claim 5, comprising a throttle valve for reducing a flow velocity of the inert gas blowing out from the gas supplying hole lower than a flow velocity of the inert gas blowing from the slit. 7. The electron beam irradiation device according to claim 5, wherein the gas supplying hole is formed as a through hole extending in a direction perpendicular to the traveling direction of the irradiated object. 8. The electron beam irradiation device according to claim 7, wherein a diameter of the gas supplying hole is greater than the gap of the slit.
	description	This application is a U.S. national stage application of International Application No. PCT/JP2006/307987, filed Apr. 17, 2006, claiming a priority date of Apr. 22, 2005, and published in a non-English language. The present invention is one relating to a working method by focused ion beam and a focused ion beam working apparatus. From olden times, there is provided a working method of repairing a photomask of semiconductor device, or the like by using a focused ion beam (Focused Ion Beam) working apparatus (for example, refer to Patent Document 1). In this disclosed working method, an ion beam emitted from an ion source is made the focused ion beam through an ion optical system (possessing a condenser lens, a blanking electrode, an aligner electrode, an aperture, an objective lens, and the like), and it is irradiated to the photomask of the semiconductor device desired to be worked (refer to FIG. 4 of the Patent Document 1). And, by blowing a gas together with this irradiation of the focused ion beam, the photomask of this semiconductor device is deposition-worked or etching-worked. In olden times, when applying a deposition working or an etching working of a predetermined pattern by using a focused ion beam apparatus, the working is implemented by dividing a region worked to the predetermined pattern into micro regions (pixels), and irradiating the focused ion beam to each of the divided pixels. And, at this time, the irradiation of the focused ion beam is performed such that a dose quantity of the focused ion beam to be irradiated becomes the same for each of the micro pixels by adjusting a scanning frequency, or the like. Patent Document 1: JP-A-2002-184342 Gazette In this working method by focused ion beam, which is provided from olden times, the following issue is left for instance. For example, as shown in FIG. 6(a), in a method of etching-working a convex part 101 of such a shape as to spread from a photomask pattern 100 toward a base by dividing it into the micro pixels like the above and scanning the focused ion beam to each of the divided pixels, there is the fact that, as shown in FIG. 6(b), one part 102 of the micro base is left while deviating from a worked area F. In order to etching-work so as to become the same face as an edge line 103 of the pattern 100 by removing this one part 102 of the base, it is demanded to irradiate the focused ion beam capable of being irradiated like a spot to this left one part 102 of the base. In other words, it is required to use the focused ion beam whose irradiation width is coincided with this left one part 102 of the base. However, there are issues that there is a technical limit in obtaining the focused ion beam having a micro beam diameter corresponding to this irradiation width, and that it is also impossible to make the irradiation width into one pixel or smaller in order to etch only this one part 102 of the base. The present invention is one made in view of the circumstances like these, and its object is to provide a working method by focused ion beam and a focused ion beam working apparatus, which can, even in a case where it is one like a opaque defect portion extending from a normal pattern in the photomask of the semiconductor device, and one whose size is smaller than the irradiation width of the focused ion beam, desirably work its place desired to be worked. As means for solving the above problems, the present invention provides a working method by focused ion beam and a focused ion beam working apparatus, which are mentioned below. A working method by focused ion beam of the present invention is a working method by focused ion beam, which performs a deposition working or an etching working to a work piece by irradiating the focused ion beam to the work piece, and characterized in that the work piece is deposition-worked or etching-worked by irradiating the focused ion beam to an edge of the work piece, and controlling a dose quantity of the focused ion beam. In the working method by focused ion beam of the present invention, the focused ion beam is irradiated to an edge made a corner part within an end part of the work piece. This edge irradiation is one meaning an irradiation irradiating the focused ion beam to the edge made the end part of a desired working place of the work piece. At this time, as the dose quantity is increased by irradiating the focused ion beam to the edge, a region to be worked gradually spreads from the edge. In other words, by controlling the dose quantity of the focused ion beam to be irradiated, its working quantity can be finely controlled. Accordingly, even in a case where the irradiation width of the focused ion beam is large, if this focused ion beam is irradiated to the edge of the work piece, it is possible to deposition-work or etching-work a region smaller than the irradiation width of the work piece with a good controllability. A working method by focused ion beam of the present invention is characterized by including a first working process performing a deposition working or an etching working to the work piece by face-irradiating the focused ion beam to an actual working range entering inside an edge part of a working range of the work piece, and a second working process performing the deposition working or the etching working to the work piece by edge-irradiating one part of the focused ion beam to an edge of an edge part or an edge part vicinity, which is left in the work piece after the first working process. In the working method by focused ion beam of the present invention, first, in the first working process, there is prepared a bit map of the actual working range entering inside the edge part of the working range of the work piece, and all of the focused ion beam is irradiated to a place coincided with this bit map and made an upper face. This face irradiation is one meaning an irradiation irradiating the focused ion beam to the upper face of a desired working place of the work piece, and one performing the deposition working or the etching working along a direction of this ion beam. In a case where the focused ion beam is irradiated like this, the deposition working or the etching working is performed along the actual working range like a normal. Incidentally, since this actual working range is made the range entering inside the edge part of the working range of the work piece, in the edge part or the edge part vicinity, there is made one in which the deposition working or the etching working is not performed intact. And, there shifts to the second working process described below. In this second working process, the deposition working or the etching working is performed to the work piece by edge-irradiating the focused ion beam to the edge of the edge part or the edge part vicinity, which is left in the work piece. This edge part means a place, in which the working finally finishes, within the desired working place of the work piece. In other words, in this second working process, the edge part becomes a final point when the deposition working or the etching working is performing by edge-irradiating. And, the inside of this edge part means the fact that it is a side reverse in direction to a working direction in the second working process. Here, in the edge part or the edge part vicinity, which is left, since the deposition working or the etching working is performed by edge-irradiating the focused ion beam, the worked face becomes one liable to be working-adjusted as mentioned above. Further, although the edge part or the edge part vicinity, which is left, is one smaller than the irradiation width of the focused ion beam, if this focused ion beam is irradiated to the edge of the work piece, it is possible to deposition-work or etching-work the edge part or the edge part vicinity, which is left, with the good controllability without decreasing this irradiation width of the focused ion beam. Further, in a case where, in a focused ion beam working apparatus performing a deposition working or an etching working to a work piece by irradiating a focused ion beam to the work piece, there is provided such a control means as mentioned above which controls so as to perform the deposition working or the etching working to the work piece by performing the deposition working or the etching working to the work piece by face-irradiating the focused ion beam to an actual working range entering inside an edge part of a working range of the work piece, and subsequently edge-irradiating the focused ion beam to an edge of an edge part left in the work piece after the first working, there can be made the focused ion beam working apparatus bringing about such actions as mentioned above. A working method by focused ion beam of the present invention is characterized in that an ion dose total quantity of the focused ion beam to be irradiated is set while corresponding to a working quantity of the edge of the work piece. Incidentally, this working quantity that the focused ion beam works the work piece is one depending on the total quantity of the ion dose contained in the focused ion beam to be irradiated to the work piece. In the working method by focused ion beam of the present invention, the ion dose total quantity of the focused ion beam to be irradiated is set in conformity with the working of the edge of the work piece, in other words, in conformity with a desired quantity with which this edge is desirably worked. By this, this edge of the workpiece is desirably deposition-worked or etching-worked by a desired quantity. More additionally, in a case where this dose quantity per unit time of the focused ion beam is decreased as the ion dose total quantity approaches to an ion dose total quantity conforming to a desired working quantity, a working place of this work piece can be finished with a better precision. A working method by focused ion beam of the present invention is characterized in that the focused ion beam is constituted by a pulse made a predetermined dose quantity, and the pulse is continued to be irradiated till becoming the ion dose total quantity. In the working method by focused ion beam of the present invention, since the focused ion beam is constituted by the pulse whose dose quantity is made the previously determined quantity, and the focused ion beam is irradiated by this pulse till becoming the ion dose total quantity in conformity with a desired quantity with which the edge is desirably worked, if a frequency of this pulse to be irradiated is controlled, this focused ion beam can be made an ion dose total quantity capable of desirably working. Accordingly, by the irradiation frequency of the pulse, since there can be adjusted to the ion dose total quantity capable of desirably working, it is possible to desirably control an irradiation quantity of the focused ion beam. More additionally, in a case where this pulse is irradiated to the work piece continuously in a commencement of the irradiation and with a suitable time interval being inserted as becoming near the ion dose total quantity, the working place of this work piece can be finished with the good precision. According to the working method by focused ion beam and the focused ion beam working apparatus, which are concerned with the present invention, even in a case where the place desired to be worked is one smaller than the irradiation width of the focused ion beam, that place desired to be worked can be worked with the good precision. Hereunder, as to embodiments about a focused ion beam working apparatus and a working method by this focused ion beam working apparatus, which are concerned with the present invention, there are explained while referring to the drawings. FIG. 1 is a schematic diagram of an ion beam working apparatus concerned with the present invention, FIG. 2 a view showing a working place by a bit map, FIG. 3 a view showing a first working process, FIG. 4 a view showing a second working process, and FIG. 5 a graph showing a relation between a dose quantity and a movement quantity (worked quantity of a work piece) of an edge. A reference numeral 1 shown in the block diagram of FIG. 1 is a focused ion beam working apparatus concerned with the present invention, and one working a work piece 90 including the photomask of the semiconductor device in which a pattern comprising Cr or MoSi is formed on a substrate comprising SiO2. This focused ion beam working apparatus 1 is constituted basically by an ion source 10 generating an ion beam I, an ion optical system 20 focussing this ion beam, and a stage 40 supporting the work piece and possessing an XY-movement mechanism. Additionally, in this focused ion beam working apparatus 1, there are provided a gas gun 45 shooting toward the work piece 90 supported by the stage 40, a charged particle detector 47 detecting a secondary charged particle 91 generating from the work piece 90, and a neutralizer 49 neutralizing the charged work piece 90. Incidentally each of the ion source 10, the ion optical system 20, the gas gun 45, the charged particle detector 47 and the neutralizer 49 is connected to a computer (control means) 50 and, by this computer 50, a suitable control is performed. The ion source 10 is a source generating the ion beam I (refer to FIG. 4), and a gallium ion made a liquid metal is used for instance. This gallium ion is easy to be designed as the ion source from a reason that its melting point is low, or the like, and desirable as a selection of the ion source 10. Additionally, in the ion optical system 20 focussing the ion beam I generating from the ion source 10, there are included a condenser lens 21 for focussing the ion beam I generated in the ion source 10, a blanking electrode 22 deflecting the beam in order that the ion beam I is not irradiated to the work piece 90, the condenser lens 21 forming an electric field to thereby focus the ion beam I, an objective lens 23 focussing the focused ion beam I to a desired working place of the work piece 90, and the like. Further, although not shown in the drawing, it has also a deflector deflection-scanning the focused ion beam. Incidentally, in the ion optical system 20, depositions of these may be suitably altered, and it may be one in which there are additionally provided suitable aligner, aperture and the like. Further, the focused ion beam I that the focused ion beam working apparatus 1 generates and irradiates to the work piece 90 is constituted by pulses, and one pulse is set to a predetermined dose quantity. The gas gun 45 is one blowing a gas toward a desired deposition working place of the work piece 90 when performing a deposition working by irradiating the focused ion beam I, and possesses a suitable nozzle, thereby blowing a suitable gas forming a deposition film. The charged particle detector 47 detects the secondary charged particle 91 generating from the work piece 90 in a case where the focused ion beam I is irradiated to and scanned on the workpiece 90. The neutralizer 49 is one neutralizing the work piece 90 charged by the fact that the focused ion beam I is irradiated, and constituted by an electron gun irradiating an electron beam, or the like. The stage 40 is one mounting and supporting the work piece 90 and, though not shown in the drawings, in it, there is provided a movement mechanism movable in an XY-direction, which constitutes a two-dimension while intersecting perpendicularly to each other. Further, the computer 50 possesses a control section, a storage section, an input section, a display section, and the like. This computer 50 displays an image of the work piece 90 on the basis of a signal from the charged particle detector 47 detecting the secondary charged particle 91 generating when the focused ion beam I is scan-irradiated to the work piece 90, besides it controls each part of the focused ion beam working apparatus 1. Further, on the basis of information of the image concerned, the computer 50 prepares a bit map M in which a pattern working region of the work piece 90 is divided into micro regions (pixels). The focused ion beam working apparatus 1 constituted like this works the work piece 90 as follows by irradiating the focused ion beam I to the work piece 90. Incidentally, on occasion of the working like this, each part constituting the above-mentioned focused ion beam working apparatus 1 becomes one controlled by the computer 50 and, in that control, there are included a first working process and a second working process. Incidentally, FIG. 2(a) and FIG. 3(a) are drawings relating to a deposition working, and FIG. 2(b) and FIG. 3(b) drawings relating to an etching working. In the first working process, first, a range entering inside an edge part of a working range of the work piece 90 is determined as an actual working range in which the working is actually performed in the first working process, thereby preparing the bit map M about this actual working range. Concretely, first, as shown in FIG. 2(a) and FIG. 2(b), the focused ion beam I is scan-irradiated to the work piece (photomask pattern) 90 by the ion source 10 and the ion optical system 20. As shown also in FIG. 1, in a case where the focused ion beam I is irradiated to the work piece 90, the secondary charged particle 91 generates from this work piece 90. This generated secondary charged particle 91 is detected by the charged particle detector 47 and, on the basis of an intensity of the detected secondary charged particle 91 of this secondary charged particle 91 and position information of the ion beam I, the computer 50 performs an imaging processing of the work piece 90. On the basis of this image information, the computer 50 prepares the bit map M of the work piece 90. That is, as shown in FIG. 2(a) and FIG. 2(b), there is formed the bit map M in which the work piece 90 is mapped in a predetermined range unit. Incidentally, a concave part 92 in FIG. 2(a) becomes a desired place in which it is desired to perform the deposition working, a convex part 93 in FIG. 2(b) becomes a desired place in which it is desired to perform the etching working, and the deposition working or the etching working is performed in order that the concave part 92 or the convex part 93 is flattened in the same plane as an edge part 96 of the work piece (photomask pattern) 90. Further, as shown in FIG. 3(a) and FIG. 3(b), there is made one in which a focused ion beam irradiation point B is irradiated for every unit range of the bit map M shown in FIG. 2. As shown in FIG. 3(a) and FIG. 3(b), a length of a spacing between the focused ion beam irradiation points becomes W1. The workpiece 90 becomes one desirably worked by the irradiation of this ion beam I. After the bit map M is prepared like this, subsequently the focused ion beam working apparatus 1 scans the focused ion beam I to each pixel position in order while coinciding with this bit map M, thereby face-irradiating each pixel position. This face irradiation is one meaning an irradiation in which the focused ion beam I is irradiated to an upper face of a desired working place of the work piece 90 such that the ion beam I is irradiated to each of the unit range of the prepared bit map M, and one performing the deposition working or the etching working along a direction of this ion beam. That is, in the deposition working in the first working process, there is made one in which the deposition film is gradually laminated by scan-irradiating the focused ion beam I to each pixel position of the concave part 92 while supplying a gas for the deposition from the gas gun 45, and buried till a slight inside in which all the concave part is buried. Further, in the etching working in the first working process, there is made one in which the convex part 93 is gradually cut by scan-irradiating the focused ion beam I, and cut till a slight inside from a boundary line of a set working region. Thus in the first working processing, the focused ion beam I is scan-irradiated in a working range that does not overlap the work piece edge C by a distance smaller than the irradiation width W1 of the focused ion beam. By doing like this, it follows that, in the deposition working, a slight, micro concave part (missing portion) 94 is left at an edge part or an edge part vicinity as shown in FIG. 3(a) and, in the etching working, a slight, micro convex part (excess portion) 95 is left at the edge part or the edge part vicinity as shown in FIG. 3(b). As shown in FIG. 4(a) and FIG. 4(b), a length of a width in a micro part of each of the micro concave part 94 and the micro convex part 95, which are left slightly, becomes W2. As shown in the drawing, this length W2 of the micro part becomes small in comparison with the length W1 of the above-mentioned spacing between the focused ion beam irradiation points, i.e., the length W2 is smaller than the irradiation width W1 of the focused ion beam I. Further, each of the micro concave part 94 and the micro convex part 95, which are left slightly, becomes a place finally worked among the desired working places of the work piece. By doing like this, the first working process is finished, and subsequently there shifts to the second working process. In this second working process, as shown in FIG. 3(a), in the deposition working, the micro concave part (missing portion) 94 left slightly in the above-mentioned first working process is additionally deposition-worked. Further, in the etching working, as shown in FIG. 3(b), the micro convex part (excess portion) 95 left slightly in the above-mentioned first working process is additionally etching-worked. Concretely, in this second working process, the focused ion beam I is edge-irradiated to an edge E made a corner part of an end part of the micro concave part 94 shown in FIG. 4(a) or the micro convex part 95 shown in FIG. 4(b). This edge irradiation is one meaning, while differing from the above-mentioned face irradiation, an irradiation irradiating the focused ion beam I to the edge E that is an end part of the desired working place of the work piece, and one performing the deposition working or the etching working by irradiating the focused ion beam I to the edge E. Incidentally, a sign C in FIG. 3(a) and FIG. 3(b), and FIG. 4(a) and FIG. 4(b) is a place corresponding to an edge part in the present invention, and this edge part C means a place in which the working finally finishes within the desired working place of the work piece 90. In other words, in this second working process, the edge part C becomes a final point in performing the deposition working or the etching working by edge-irradiating. Further, the focused ion beam I irradiated in this second working process becomes one whose dose quantity is controlled. That is, there becomes one whose dose quantity can be arbitrarily selected in conformity with the micro concave part 94 or the micro convex part 95, which is left in the above-mentioned first working process. This dose quantity is one determined by a product of an electric current quantity and an irradiation time of the focused ion beam, and one deduced as a physical quantity corresponding to the irradiation time if the electric current quantity of the focused ion beam is stable one. Incidentally, this focused ion beam I may be constituted by pulses set to a predetermined electric current quantity of the focused ion beam. In this case, the dose quantity to be irradiated becomes one determined by a frequency of the pulses to be irradiated. And, a working quantity of the deposition working or the etching working becomes one depending on the dose quantity of the focused ion beam I to be irradiated. In other words, as shown in a graph of FIG. 5, which shows a relation between the dose quantity and a movement quantity of the edge (in other words, a quantity with which a position of the edge moves by the fact that a region gradually worked spreads from the edge), as the dose quantity increases also the movement quantity of the worked edge increases. Concretely, in a case where the dose quantity of the focused ion beam I is 400 CST (CST: a certain unit proportional to the dose quantity), the movement quantity of the edge becomes 3.5 nm and, in a case where the dose quantity of the focused ion be an B is 500 CST, the movement quantity of the edge becomes one near 11 nm. Incidentally, this movement quantity of the edge is made the same as the working quantity that the focused ion beam I works. In the second working process, since the dose quantity in regard to the movement quantity of the edge is previously measured like this FIG. 5, in a case working this edge, it is possible to irradiate the focused ion beam I corresponding to a dose quantity for obtaining a desired edge movement quantity. In other words, it is possible to arbitrarily select and set the dose quantity in view of the above-mentioned working quantity of the micro concave part 94 or the micro convex part 95 and, by this, the working can be performed with a good precision. The focused ion beam I whose dose quantity is selected and set like this is irradiated to the edge E near the edge part C in regard to the work piece 90 as shown in FIG. 4(a) and FIG. 4(b). By doing so, in the deposition working of FIG. 4(a), the deposition film is gradually formed in this edge E, and this micro concave part 94 becomes buried one. In other words, by this deposition working, the deposition film is gradually formed in the edge E toward a right in the drawing, and this micro concave part 94 is buried. Further, in the etching working of FIG. 4(b), this edge E is gradually cut, and this micro convex part 95 becomes removed one. In other words, by this etching working, the edge E is gradually cut toward a left in the drawing, and this micro concave part 94 is removed. By doing like this, even in a case where a place, in which it is desired to perform the working, is one whose width is smaller than the irradiation width of the focused ion beam I, it is possible to desirably work the place in which it is desired to perform the working. Incidentally, a technical scope of the present invention is not one limited to the above embodiment, and it is possible to add various modifications in a scope not deviating from a gist of the present invention. For example, in the above-mentioned embodiment, although the focused ion beam is constituted by the pulses made the predetermined dose quantity, there is not limited to this, and it is also possible to continuously irradiate the focused ion beam and adjust the dose quantity by its irradiation time. Further, although the gallium ion is used as the ion source, there is not limited to this, and it is possible to use a suitable ion source. Further, also as to the work piece, it is not one limited to the photomask, and it is possible to work a suitable work piece constituted very small. Even in the case where it is one like the opaque defect portion extending from the normal pattern in the photomask of the semiconductor device, and one whose size is smaller than the irradiation width of the focused ion beam, it is possible to desirably work its place desired to be worked.
048246344	claims	1. In a fuel element for use in a nuclear reactor which includes a fission material contained within a zirconium-alloy cladding tube, the improvement which comprises: a coating on the inside of the zirconium-alloy cladding tube, said coating including boron-containing compound burnable poison particles deposited from a liquid suspension which includes an acrylic polymer binder material. 2. The fuel element of claim 1 in which the coating's boron-containing compound includes boron enriched to at least an 80% level of B.sup.10 to give a desired nuclear poison level for use in the nuclear reactor. 3. The fuel element of claim 1 in which the coating's boron-containing compound includes zirconium diboride. 4. The fuel element of claim 1 in which the zirconium-alloy includes tin in the approximate range of from 1.20 to 1.70. 5. The fuel element of claim 1 in which the particles range in size below 1.5 microns in a distribution in which 20% of the particles are of a size greater than 1 micron and 80% of the particles are of a size less than 1 micron. 6. The fuel element of claim 1 in which the coating includes residual acrylic polymer binder material. 7. The fuel element of claim 1 in which the coating includes particles of graphite as well as particles of a boron-containing compound deposited from the liquid suspension which includes the acrylic polymer binder.
	claims	1. A system for storing heat generated in a nuclear power plant, the system comprising at least one tank;wherein the at least one tank contains or is filled with high temperature resistant solids; the high temperature resistant solids comprising at least one of particles and pebbles, the particles and pebbles made of at least one of alumina, silica, quartz, and ceramic; the at least one tank being configured to receive a first fluid in order to transfer and store heat from the first fluid to the high temperature resistant solids; the at least one tank being also configured to receive a second fluid in order to transfer the heat from the high temperature resistant solids to the second fluid;wherein the at least one tank and the high temperature resistant solids are configured such that heat is absorbed in the high temperature resistant solids and that the heat spreads through the at least one tank in a relatively sharp front; andwherein the relatively sharp front has a width that is less than one tenth of a length of the at least one tank. 2. The system of claim 1, wherein the first fluid comprises a compressed gas. 3. The system of claim 2, wherein the compressed gas comprises helium. 4. The system of claim 1, wherein the second fluid comprises a compressed gas. 5. The system of claim 1, wherein the first fluid comprises a compressed gas moving at a predetermined velocity. 6. The system of claim 1, wherein the first fluid is higher in temperature than the second fluid. 7. The system of claim 1, wherein the first fluid passes through at least one device heated by nuclear fission before entering the at least one tank. 8. The system of claim 1, wherein the second fluid is used to produce steam in a power plant before entering the at least one tank. 9. The system of claim 1, wherein the first fluid comprises a compressed gas passing through at least one nuclear reactor core. 10. The system of claim 1, wherein the second fluid comprises a compressed gas passing through a power plant generating electrical power. 11. The system of claim 1, wherein the solid media comprises:at least one packed bed of at least one of particles and pebbles. 12. The system of claim 1, wherein the system is structured and arranged to move at least one of the first and second fluids through the at least one tank with at least one of uniform flow distribution and minimal pressure drops. 13. The system of claim 1, wherein the first and second fluids comprise portions of the same compressed gas flowing in a closed system, wherein the portions have different temperatures when entering the at least one tank. 14. The system of claim 1, wherein the first fluid comprises a fluid heated by at least one reactor core before entering the at least one tank and the second fluid comprises a fluid exiting a power plant before entering the at least one tank. 15. A system for storing heat generated in a nuclear power plant, the system comprising:at least one tank, wherein said at least one tank contains or is filled with high temperature resistant solids; the high temperature resistant solids comprising at least one of particles and pebbles made of at least one of alumina, silica, quartz and ceramic; the at least one tank being configured to receive a first fluid in order to transfer and store heat from the first fluid to the solid media high temperature resistant solids; the at least one tank being also configured to receive a second fluid in order to transfer the heat from the solid media high temperature resistant solids to the second fluid; anda control system controlling at least one of:when the first fluid is allowed to pass through the at least one tank; andwhen the second fluid is allowed to pass through the at least one tank;wherein the at least one tank and the high temperature resistant solids are configured such that heat is absorbed in the high temperature resistant solids and that the heat spreads through the at least one tank in a relatively sharp front; andwherein the relatively sharp front has a width that is less than one tenth of a length of the at least one tank. 16. A system for storing heat generated in a nuclear power plant, the system comprising:at least one tank, wherein said at least one tank contains or is filled with high temperature resistant solids; the high temperature resistant solids comprising at least one of particles and pebbles, the particles and pebbles made of at least one of alumina, silica, quartz, and ceramic; the at least one tank being configured to receive a first fluid in order to transfer and store heat from the first fluid to the solid media high temperature resistant solid; the at least one tank being also configured to receive a second fluid in order to transfer the heat from the solid media high temperature resistant solids to the second fluid; anda control system controlling at least one of:when the first fluid is allowed to pass through the at least one tank;when the first fluid is allowed to bypass the at least one tank;when the second fluid is allowed to pass through the at least one tank; andwhen the second fluid is allowed to bypass the at least one tank;wherein the at least one tank and the high temperature resistant solids are configured such that heat is absorbed in the high temperature resistant solids and that the heat spreads through the at least one tank in a relatively sharp front wherein the relatively sharp front has a width that is less than one tenth of a length of the at least one tank. 17. A system for storing heat generated in a nuclear power plant, the system comprising:at least one tank, wherein said at least one tank contains or is filled with high temperature resistant solids; the at least one tank being configured to receive a first fluid in order to transfer and store heat from the first fluid to the solid media high temperature resistant solids; the high temperature resistant solids comprising at least one of particles and pebbles, the particles and pebbles made of at least one of alumina, silica, quartz, and ceramic; the at least one tank being also configured to receive a second fluid in order to transfer the heat from the solid media high temperature resistant solids to the second fluid;at least one nuclear reactor core heating the first fluid before the first fluid enters the at least one tank;a steam power plant receiving the heated fluid from the at least one nuclear reactor core under certain conditions and receiving the second fluid from the at least one tank under certain other conditions; andone or more valves controlling movement of the first and second fluids between the at least one nuclear reactor core, the at least one tank, and the steam power plant; and one or more recycle compressors pressurizing the first and second fluids;wherein the at least one tank and the high temperature resistant solids are configured such that heat is absorbed in the high temperature resistant solids and that the heat spreads through the at least one tank in a relatively sharp front; wherein the relatively sharp front has a width that is less than one tenth of a length of the at least one tank. 18. The system of claim 17, wherein the first and the second fluid comprises helium. 19. A system for storing heat generated in a nuclear power plant, the system comprising:at least one tank, wherein said at least one tank contains or is filled with high temperature resistant solids; the high temperature resistant solids comprising at least one of particles and pebbles, the particles and pebbles made of at least one of alumina, silica, quartz, and ceramic; the at least one tank being configured to receive a first fluid in order to transfer and store heat from the first fluid to the solid media high temperature resistant solids; the at least one tank being also configured to receive a second fluid in order to transfer the heat from the solid media high temperature resistant solids to the second fluid; wherein the at least one tank and the high temperature resistant solids are configured such that heat is absorbed in the high temperature resistant solids and that the heat spreads through the at least one tank in a relatively sharp front; wherein the relatively sharp front has a width that is less than one tenth of a length of the at least one tank;wherein the system has the following three cycles:a first cycle wherein the first fluid bypasses the at least one tank, flows to a power plant, and returns to at least one reactor core;a second cycle wherein at least a portion of the first fluid flows through the at least one tank and returns to the at least one reactor core; anda third cycle wherein the second fluid passes through the at least one tank, flows to the power plant, and returns to the at least one tank.
047909765	abstract	Device for flushing out a lance-housing tube in a reactor pressure vessel of a boiling-water reactor and for aligning therein a dry LVD lance which partly protrudes with a pressure-tight lance passthrough from an end flange on the lance-housing tube of the reactor pressure vessel, includes a tubular housing surrounding from below a part of the lance protruding from the reactor pressure vessel and sealed by a lance protection tube, the tubular housing being fastenable to the end flange; and a piston arranged in the tubular housing underneath the sealed lance, the piston being vertically displaceable and rotatable.
051130785	abstract	A radiation shielding structure including a radiation shielding panel which comprises a lead transparent plate, for example, a transparent lead acrylic resin plate, lead glass plate, etc., and a thin nonlead transparent plate, for example, a transparent acrylate resin plate, glass plate, etc., which is laminated on at least one side of the lead transparent plate. Thus, lead that is contained in the lead transparent plate shields radioactive rays, while the nonlead transparent plate, which is laminated on at least one side of the lead transparent plate, prevents oxidation of the lead in the lead transparent plate by air or chemicals, which oxidation would otherwise form an oxide film on the panel surface and make the panel opaque.
055132316	summary	TECHNICAL FIELD OF THE INVENTION The present invention relates to containers used for transportation and short term storage of spent nuclear fuel. BACKGROUND OF THE INVENTION As the nuclear utility industry matures, there is an ever-increasing need for additional storage space to safely contain spent nuclear fuel. One method that has been developed in recent years for storage of spent nuclear fuels is dry storage in horizontal storage modules, which are shielded bunkers in which containerized spent fuel is stored and monitored for definite periods of time. One conventional technique for horizontal modular dry storage of spent nuclear fuel rods is disclosed in U.S. Pat. No. 4,780,269 to Fischer et al. A basic procedure for dry storage of spent nuclear fuel is to position a dry shielded canister into a shielded transfer cask. The canister and cask are filled with deionized water, which is then lowered into a pool containing the spent nuclear fuel. Spent fuel assemblies are then placed into the canister, and a shielded end plug is positioned to close the canister. The canister and cask are then removed from the pool, and the cask and canister are drained and dried. The exterior of the cask is decontaminated, followed by closure of the cask with a closure plate. The closed transportation cask is then lowered onto a transport trailer and secured by tie-downs. The transport trailer carries the cask to the sight of the horizontal dry storage modules. The cask is opened and docked with an entry port of a dry storage module. The canister is then transferred from the cask into the module, such as by passing a ram through the dry storage module from an end opposite the entry port, through the entry port and into the opened cask. The canister can then be grasped and pulled into the dry storage module, after which both the entry port and access port are sealed. A critical aspect of this process is the safe containment and transfer of the spent nuclear fuel within the canister from the original pool storage to the final dry horizontal storage site. The transport cask must be constructed with adequate structural strength and shielding to both physically protect the dry shielded canister within, and to provide biological shielding to minimize personnel radiation dosages during canister transfer and transport operations. During the canister transfer and transport process, the cask must be able to withstand any foreseeable impact, such as could occur by accidental dropping of the cask from the transport trailer or exposure to tornadoes or other natural disasters. In the United States, federal regulations setting forth requirements that transport casks must meet are found in 10 C.F.R. 72, including subpart G, as well as 10 C.F.R. 71 and 10 C.F.R. 50. In particular, the cask must be able to withstand impacts due to a drop of 30' onto an essentially, unyielding flat horizontal surface, without structural failure. Even if structurally damaged, no leakage of the contents from the cask is permitted. It is thus important to design casks with high structural integrity. At the same time, it is desirable to maximize the quantity of spent fuel that can be transported within the cask at any given time, and to minimize the cost of constructing the cask. While strength considerations typically warrant constructing the cask from thicker sections of metal and other materials, this requirement may reduce the quantity of spent fuel that can be transported within the cask. External dimensions of the cask are limited by constraints such as the total weight of the loaded cask, and clearances required to transport the casks through tunnels, under bridges and overpasses, and the like. Currently, conventional casks are often constructed from a polished austentitic stainless steel, such as 304 stainless steel, for corrosion prevention. However, such stainless steel is limited in strength and may fail under high stresses. To combat this potential, conventional casks are constructed from thick metal sections, and must be reinforced with gusset plates and other reinforcing members. Additionally, locations on the casks that are subjected to force during transport must be reinforced with additional metal plates welded to the cask structure. For example, conventional casks are outfitted with cylindrical trunnions welded or bolted directly to a cylindrical structural shell of the cask at diametrically opposed locations. These trunnions are grasped by hooks, and serve as pivot points while lieting the cask during the transportation process. Because of the stresses transferred to the cask structure from the trunnions during use, the shell is typically reinforced in the area surrounding the trunnions by welding additional plates of metal. The trunnions themselves are conventionally permanently secured to the structural shell of casks by welding or bolting directly to the shell. In the case of welding, the welded joint is subjected to substantial stress during hoisting of the cask. In the case of bolting the trunnions in place, the bolts are subjected to extreme shear and tensile loads during hoisting of the cask. Again, the trunnions must be heavily reinforced to withstand such loads, increasing the weight and overall dimensions of the cask, and thus decreasing the spent fuel containment capacity and increasing the cost of manufacture. When sealed joints, such as elastomeric (e.g., O-ring) seals or metal seals are utilized, the base metal used to form the structural shell is conventionally machined to form the sealing surfaces. Thus, for example, when 304 stainless steel is used to construct the shell, annular surfaces on the shell are machined and polished to form sealing surfaces. While functioning adequately in most situations, extreme impact to the seal area, such as by accidental dropping of the cask at an oblique angle whereby force is concentrated on the seal area, may result in permanent deformation of the metal seal surface, and subsequent leakage potential. SUMMARY OF THE INVENTION The present invention provides a container designed for use as a cask for short-term containment and transporting of spent nuclear fuel. In the first aspect of the present invention, the container is formed from a structural shell defining a cavity for receiving spent nuclear fuel, and first and second end apertures opening into the cavity. The shell has a first end portion formed of a first material and a second end portion formed of a second material. The first end portion is joined to the second end portion to form the structural shell. A bearing surface is defined on the first end portion of the shell and is engageable to enable hoisting of the container. The first end portion of the shell is constructed from a first material that has a higher load bearing strength than the second material, to handle the hoisting stress. The container also includes a first closure securable to the first end portion of the shell to seal the first end aperture, and a second closure securable to the second end portion of the shell to seal the second end aperture. The container further includes a radiation absorbing shield layer, which may include both gamma radiation and a neutron radiation absorbing materials. The container is thus constructed so that those areas of the container that are subjected to the greatest stress, e.g. the first end portion, is constructed from the strongest material, such as a high-strength metal alloy. However, those portions of the cask that are not exposed to as high a stress are produced from lower cost materials having a strength that is adequate for the lower loads to be imposed on those portions. In a further aspect of the present invention, a cask is provided that includes a tubular inner shell defining a cavity for receiving spent nuclear fuel, and first and second ends. A tubular outer shell having first and second ends is assembled coaxially over the inner shell to define an annular space therebetween. A radiation absorbing material fills the annular space. An annular member defining a central aperture and a first annular sealing surface is secured about its perimeter to the first ends of the inner shell and the outer shell to create airtight joints with both the inner shell and the outer shell. A first closure plate is releasably securable to the annular member and defines a second annular sealing surface corresponding to the first annular sealing surface defined by the annular member. A seal is positioned between the second annular sealing surface of the first closure plate and the first annular sealing surface of the annular member to create an airtight seal between the first closure plate and the annular member. The cask also includes a second closure plate secured proximate its perimeter to the second ends of the inner shell to create airtight joints with the inner shell. In a further aspect of the present invention, a cask is provided that includes a structural shell defining a cavity for receiving spent nuclear fuel and first and second end apertures. A first closure is securable to the shell to seal the first end aperture. A second closure is securable to the shell to seal the second end aperture. A radiation absorbing shield layer is affixed to the shell. First and second pairs of trunnion mounting structures, preferably configured as tubular sleeves are secured in opposing disposition within apertures formed in the structural shell. First and second trunnions, each defining a base and a beating surface, are included. The base of each trunnion is releasably securable to a corresponding one of the trunnion mounting structures, whereby the bearing surfaces of the first trunnions can be grasped to hoist the container. The second trunnions are used to provide a point of support and rotation for loading and unloading the cask from its conveyance. In a preferred embodiment, the trunnion mounting structures are configured as annular sleeves that are welded to the structural shell of the cask, within which sleeves the base of the trunnions are received. Because of this construction, fasteners such as bolts used to secure the trunnions to the mounting structures are substantially isolated from tensile and shear loads. In a further aspect of the present invention, the trunnion mounting structures are preferably formed from a high-strength material such as is used to form the portion of the outer shell to which the first trunnions are mounted, thereby providing a strong trunnion mounting without requiting additional plate reinforcement. In a still further aspect of the present invention, improved seal joints are included in the cask. Sealing surfaces of the cask are formed utilizing hardened metal weld overlays, thereby providing sealing surfaces that are not readily subject to permanent deformation upon impact of the cask. In the preferred embodiment, sealing surfaces of closure plates on the cask include grooves formed to define a half-dovetailed cross section for receiving seals. This enables use of either metal or elastomeric seals in the joints, and enables assembly of the joints while the cask is in either the horizontal or vertical disposition. In a still further aspect of the present invention, a cask is disclosed that includes a tubular structural shell defining a cavity for receiving spent nuclear fuel and first and second opened ends. The first closure plate is releasably securable to the first opened end of the shell, whereby when secured to the shell, the first opened end of the shell is sealed, and when released from the shell, loading and unloading of spent nuclear fuel through the first open end into the cavity is permitted. The second closure plate is secured to and seals the second open end of the shell. The second closure plate defines a central access aperture. An access cover plate is releasably securable to the second closure plate to seal the central access aperture. When released from the second closure plate, entry of a ram through the access aperture into the cavity of the shell to facilitate unloading of spent nuclear fuel through the first open end of the shell is permitted. Shield plugs filled with a radiation-absorbing material are provided to cover the trunnion mounting structures and central access aperture formed in the cask during short-term storage and transportation. In another aspect, the present invention relates to a skid for transporting a nuclear fuel transportation cask and containment vessel. The skid supports the cask around the neutron radiation shielding material. The skid includes a supporting member and a retaining member that each include a plurality of parallel spaced-apart plates lying in planes perpendicular to a longitudinal axis of the cask which are connected by a plurality of longitudinal fins parallel to the longitudinal axis of the cask. The longitudinal fins are positioned to mate with structural elements associated with the neutron radiation shielding material to transfer loads from the cask to the skid. The present invention thus provides a cask that is less costly to construct, yet that provides improved safety under impact conditions. Exposure of workers to radiation during transport procedures is also reduced.
	claims	1. A pattern generation system which generates beam control data for causing at least one beam to image a pattern on a target, said pattern generation system comprising: a data generation subsystem configured so that a set of hierarchical image data having at least three levels of hierarchy is generated and asserted, wherein the hierarchical image data includes residual data and a set of cells, each of the cells determines a feature set of the pattern, the residual data includes at least two subroutine call commands, and each of the subroutine call commands identifies a cell of the set of cells and a portion of the target at which the feature set determined by the cell is to be imaged; and a graphics engine having a memory, wherein the graphics engine is coupled and configured so that hierarchical image data is received, and said set of cells is stored upon receipt of the hierarchical image data, so that beam control data is generated in response to the residual data and the contents of the memory, wherein the graphics engine is configured so that a response is made to each of at least two of the subroutine call commands by retrieving one of the cells from the memory and generating a portion of the beam control data which controls imaging of the feature set determined by said one of the cells at the portion of the target identified by each of at least two of the subroutine call commands, wherein the set of cells includes at least one primary cell and at least one secondary cell, the subroutine call commands include at least two primary call commands and at least two secondary call commands, the primary cell includes said at least two secondary call commands, the graphics engine is configured so that a response is made to each of at least two of the primary call commands by retrieving the primary cell from the memory and generating a portion of the beam control data which controls imaging of the feature set determined by the primary cell at the portion of the target identified by said each of at least two of the primary call commands, and the graphics engine is configured so that a response is made to each of at least two of the secondary call commands by retrieving the secondary cell from the memory and generating a portion of the beam control data which controls imaging of the feature set determined by the secondary cell at the portion of the target identified by each of at least two of the secondary call commands. 2. The system of claim 1 , wherein the beam control data is in pixel format. claim 1 3. The system of claim 1 , also including: claim 1 a beam system, coupled to receive the beam control data from the graphics engine and configured so that the pattern on the target is imaged in response to the beam control data, wherein the beam system is configured so that a response is made to the beam control data by operating in a sequence of configurations which cause the at least one beam to image the pattern on the target. 4. The system of claim 3 , wherein the graphics engine is a raster engine, and wherein the beam system is configured so that a response is made to the beam control data by executing a raster scan of the at least one beam relative to the target. claim 3 5. The system of claim 1 , wherein the data generation subsystem includes a processor programmed to generate the set of hierarchical image data in response to a set of hierarchical raw image data, wherein the processor is programmed with software for: claim 1 determining a hierarchy graph for the raw image data, identifying said hierarchy graph as a tentative hierarchy graph of a tentative version of the hierarchical image data, sorting cells of the tentative hierarchy graph according to a figure of merit, and classifying the cells of the tentative hierarchy graph into a first cell set and a second cell set such that each cell in the first cell set satisfies at least one predetermined criterion including the criterion that the figure of merit of each cell of the first cell set has a predetermined relation to a threshold value, and second cell set consists of each of the cells that is not in the first cell set; and determining the hierarchical image data from the hierarchical raw image data by replacing the second cell set and the subroutine calls to each cell in the second cell set with replacement residual data, such that the hierarchy graph of the hierarchical image data includes the first cell set but not the second cell set, and the residual data of the hierarchical image data includes the replacement residual data. 6. The system of claim 1 , wherein the data generation subsystem includes a processor programmed to generate the set of hierarchical image data in response to a set of hierarchical raw image data, wherein the processor is programmed with software for: claim 1 determining a hierarchy graph for the raw image data, identifying said hierarchy graph as a tentative hierarchy graph of a tentative version of the hierarchical image data, and sorting cells of the tentative hierarchy graph according to their individual size relative to total size of the set of raw image data; identifying one of the cells of the tentative hierarchy graph whose size is largest relative to the total size of the tentative version of the hierarchical image data, and if the size of said one of the cells does not exceed a cachable size, determining whether inclusion of said one of the cells as a cell of the hierarchical image data would cause the size of the residual portion of the hierarchical image data to exceed a predetermined maximum size; and if inclusion of said one of the cells as a cell of the hierarchical image data would cause the size of the residual portion of the hierarchical image data to exceed the predetermined maximum size, identifying a new cell that contains multiple instantiations of said one of the cells and has size that does not exceed the cachable size, updating the tentative hierarchy graph and the tentative version of the hierarchical image data by replacing the multiple instantiations of said one of the cells with said new cell, thereby determining an updated hierarchy graph and an updated version of the hierarchical image data, and if the size of the residual portion of the updated hierarchical image data does not exceed the predetermined maximum size, identifying the updated hierarchical image data as the hierarchical image data. 7. A graphics engine which generates beam control data for causing at least one beam to image a pattern on a target, said graphics engine comprising: a memory coupled to receive a set of hierarchical image data having at least three levels of hierarchy, wherein the hierarchical image data includes residual data and a set of cells, each of the cells determines a feature set of the pattern, the residual data includes at least two subroutine call commands, and each of the subroutine call commands identifies a cell of the set of cells and a portion of the target at which the feature set determined by the cell is to be imaged, and wherein the engine is configured so that said set of cells is stored in the memory upon receipt of the hierarchical image data; and a beam control data generator coupled to the memory, wherein the beam control data generator is coupled to receive at least the residual data of the hierarchical image data and configured so that beam control data is generated in response to the residual data and the contents of the memory, and wherein the beam control data generator is configured so that a response is made to each of at least two of the subroutine call commands by retrieving one of the cells from the memory and generating a portion of the beam control data which controls imaging of the feature set determined by said one of the cells at the portion of the target identified by each of at least two of the subroutine call commands, wherein the set of cells includes at least one primary cell and at least one secondary cell, the subroutine call commands include at least two primary call conunands and at least two secondary call commands, the primary cell includes said at least two secondary call commands, the beam control data generator is configured so that a response is made to each of at least two of the primary call commands by retrieving the primary cell from the memory and generating a portion of the beam control data which controls imaging of the feature set determined by the primary cell at the portion of the target identified by said each of the primary call commands, and the beam control data generator is configured to respond to each of at least two of the secondary call commands by retrieving the secondary cell from the memory and generating a portion of the beam control data which controls imaging of the feature set determined by the secondary cell at the portion of the target identified by said each of at least two of the secondary call commands. 8. The engine of claim 7 , wherein the beam control data is in pixel format. claim 7 9. A pattern generation system which generates beam control data for causing at least one beam to image a pattern on a target, said pattern generation system comprising: a data generation subsystem configured so that a set of hierarchical image data having at least three levels of hierarchy is generated and asserted, wherein the hierarchical image data includes residual data and cells, the cells include at least one primary cell determining a feature set of the pattern and at least one secondary cell determining a feature set of the pattern, the residual data includes at least two subroutine call commands for each of the cells, and each of the subroutine call commands identifies one of the cells and a portion of the target at which the feature set determined by said one of the cells is to be imaged; and a graphics engine having a memory, wherein the graphics engine is coupled and configured so that the hierarchical image data is received, and in response, a beam control data cell is generated and stored in the memory in response to each of the cells of the hierarchical image data, and beam control data is generated in response to the residual data and the contents of the memory, and wherein the graphics engine is configured so that a response is made to each of the subroutine call commands by retrieving one said beam control data cell from the memory. 10. The system of claim 9 , wherein the beam control data is in pixel format. claim 9 11. The system of claim 9 , also including: claim 9 a beam system, coupled to receive the beam control data from the graphics engine and configured so that the pattern on the target is imaged in response to the beam control data, wherein the beam system is configured so that a response is made to the beam control data by operating in a sequence of configurations which cause the at least one beam to image the pattern on the target. 12. The system of claim 11 , wherein the graphics engine is a raster engine, and wherein the beam system is configured so that a response is made to the beam control data by executing a raster scan of the at least one beam relative to the target. claim 11 13. A graphics engine which generates beam control data for causing at least one beam to image a pattern on a target, said graphics engine comprising: a memory; and a beam control data generator coupled to the memory and to receive a set of hierarchical image data having at least three levels of hierarchy, wherein the hierarchical image data includes residual data and cells, the cells include at least one primary cell determining a feature set of the pattern and at least one secondary cell determining a feature set of the pattern, the residual data includes at least two subroutine call commands for each of the cells, and each of the subroutine call commands identifies one of the cells and a portion of the target at which the feature set determined by said one of the cells is to be imaged, wherein the beam control data generator is configured so that a bema control data cell is generated and stored in the memory in response to each of the cells, and so that the beam control data is generated in response to the residual data and the contents of the memory, and wherein the beam control data generator is configured so that a response is made to each of the at least two of the subroutine call commands by retrieving one said beam control data cell from the memory. 14. The engine of claim 13 , wherein the beam control data is in pixel format. claim 13 15. A processor programmed to generate a set of hierarchical image data having at least two levels of hierarchy in response to a set of hierarchical raw image data, such that the hierarchical image data includes residual data and at least one cell, each said cell determines a feature set of a pattern, the residual data includes at least two subroutine call commands, and each of the subroutine call commands identifies a cell of said at least one cell and a portion of a target at which the feature set determined by the cell is to be imaged, wherein the processor is programmed with software for: (a) determining a hierarchy graph for the raw image data, identifying said hierarchy graph as a tentative hierarchy graph of a tentative version of the hierarchical image data, sorting cells of the tentative hierarchy graph according to a figure of merit, and classifying the cells of the tentative hierarchy graph into a first cell set and a second cell set such that each cell in the first cell set satisfies at least one predetermined criterion including the criterion that the figure of merit of each cell of the first cell set has a predetermined relation to a threshold value, and second cell set consists of each of the cells that is not in the first cell set; and (b) determining the hierarchical image data from the hierarchical raw image data by replacing the second cell set and the subroutine calls to each cell in the second cell set with replacement residual data, such that the hierarchy graph of the hierarchical image data includes the first cell set but not the second cell set, and the residual data of the hierarchical image data includes the replacement residual data. 16. The processor of claim 15 , wherein said processor is programmed with software for classifying the cells of the tentative hierarchy graph into the first cell set and the second cell set such that the figure of merit of each cell of the first cell set exceeds the threshold value and the figure of merit of each cell of the second cell set does not exceed the threshold value. claim 15 17. The processor of claim 16 , wherein the cells of the hierarchical image data are to be cached in a memory of a graphics engine, and the figure of merit for each cell is the size of that portion of the memory that would be occupied by said cell if said cell were cached in said memory. claim 16 18. The processor of claim 16 , wherein the hierarchical image data is to be transferred to the graphics engine, the cells of the hierarchical image data are to be cached in a memory of the graphics engine, and the figure of merit for each cell is SAV/SIZE, where SIZE is the size of that portion of the memory that would be occupied by the cell if said cell were cached in said memory, and SAV is the data volume by which the overall volume of the hierarchical image data is reduced by including said cell in the hierarchical image data, along with subroutine call commands for repeatedly calling said cell after it has been cached in the memory, in place of residual data corresponding to said cell. claim 16 19. A processor programmed to generate a set of hierarchical image data having at least two levels of hierarchy in response to a set of hierarchical raw image data, such that the hierarchical image data includes residual data and a set of cells, each of the cells determines a feature set of a pattern, the residual data includes at least two subroutine call commands, and each of the subroutine call commands identifies a cell of the set of cells and a portion of a target at which the feature set determined by the cell is to be imaged, wherein the processor is programmed with software for: (a) determining a hierarchy graph for the raw image data, identifying said hierarchy graph as a tentative hierarchy graph of a tentative version of the hierarchical image data, and sorting cells of the tentative hierarchy graph according to their individual size relative to total size of the set of raw image data; (b) identifying one of the cells of the tentative hierarchy graph whose size is largest relative to the total size of the tentative version of the hierarchical image data, and if the size of said one of the cells does not exceed a cachable size, determining whether inclusion of said one of the cells as a cell of the hierarchical image data would cause the size of the residual portion of the hierarchical image data to exceed a predetermined maximum size; and (c) if inclusion of said one of the cells as a cell of the hierarchical image data would cause the size of the residual portion of the hierarchical image data to exceed the predetermined maximum size, identifying a new cell that contains multiple instantiations of said one of the cells and has size that does not exceed the cachable size, updating the tentative hierarchy graph and the tentative version of the hierarchical image data by replacing the multiple instantiations of said one of the cells with said new cell, thereby determining an updated hierarchy graph and an updated version of the hierarchical image data, and if the size of the residual portion of the updated hierarchical image data does not exceed the predetermined maximum size, identifying the updated hierarchical image data as the hierarchical image data. 20. A method for generating a set of hierarchical image data having at least two levels of hierarchy in response to a set of hierarchical raw image data, such that the hierarchical image data includes residual data and a set of cells, each of the cells determines a feature set of a pattern, the residual data includes at least two subroutine call commands, and each of the subroutine call commands identifies a cell of the set of cells and a portion of a target at which the feature set determined by the cell is to be imaged, said method including the steps of: (a) determining a hierarchy graph for the raw image data, identifying the hierarchy graph as a tentative hierarchy graph of a tentative version of the hierarchical image data, and sorting cells of the tentative hierarchy graph according to their individual size relative to total size of the set of raw image data; (b) identifying one of the cells of the tentative hierarchy graph whose size is largest relative to the total size of the tentative version of the hierarchical image data, and if the size of said one of the cells does not exceed a cachable size, determining whether inclusion of said one of the cells as a cell of the hierarchical image data would cause the size of the residual portion of the hierarchical image data to exceed a predetermined maximum size; and (c) if inclusion of said one of the cells as a cell of the hierarchical image data would cause the size of the residual portion of the hierarchical image data to exceed the predetermined maximum size, identifying a new cell that contains multiple instantiations of said one of the cells and has size that does not exceed the cachable size, updating the tentative hierarchy graph and the tentative version of the hierarchical image data by replacing the multiple instantiations of said one of the cells with said new cell, thereby determining an updated hierarchy graph and an updated version of the hierarchical image data, and if the size of the residual portion of the updated hierarchical image data does not exceed the predetermined maximum size, identifying the updated hierarchical image data as the hierarchical image data. 21. A method for generating a set of hierarchical image data to be transferred to a graphics engine, in response to a set of hierarchical raw image data, such that the hierarchical image data has at least two levels of hierarchy and includes residual data and a set of cells, each of the cells determines a feature set of a pattern, the residual data includes at least two subroutine call commands, and each of the subroutine call commands identifies a cell of the set of cells and a portion of a target at which the feature set determined by the cell is to be imaged, said method including the steps of: determining a hierarchy graph for the raw image data, and determining a tentative version of the hierarchical image data having said hierarchy graph; and modifying the tentative version of the hierarchical image data to determine an optimized version of the hierarchical image data which achieves an optimal combination of reduced data volume relative to the data volume of the raw image data, and reduced time required for generation of beam control data from the optimized hierarchical image data in the graphics engine relative to the time required for generation of the beam control data from the raw image data in the graphics engine. 22. A method for generating a set of hierarchical image data having at least two levels of hierarchy in response to a set of hierarchical raw image data, such that the hierarchical image data includes residual data and at least one cell, each said cell determines a feature set of a pattern, the residual data includes at least two subroutine cell commands, and each of the subroutine call commands identifies a cell of said at least one cell and a portion of a target at which the feature set determined by the cell is to be imaged, said method including the steps of: (a) determining a hierarchy graph for the raw image data, identifying said hierarchy graph as a tentative hierarchy graph of a tentative version of the hierarchical image data, sorting cells of the tentative hierarchy graph according to a figure of merit, and classifying the cells of the tentative hierarchy graph into a first cell set and a second cell set such that each cell in the first cell set satisfies at least one predetermined criterion including the criterion that the figure of merit of each cell of the first cell set has a predetermined relation to a threshold value, and second cell set consists of each of the cells that is not in the first cell set; and (b) determining the hierarchical image data from the hierarchical raw image data by replacing the second cell set and the subroutine calls to each cell in the second cell set with replacement residual data, such that the hierarchy graph of the hierarchical image data includes the first cell set but not the second cell set, and the residual data of the hierarchical image data includes the replacement residual data. 23. A method for generating beam control data for causing at least one beam to image a pattern on a target, said method comprising the steps of: (a) generating a set of hierarchical image data having at least three levels of hierarchy, wherein the hierarchical image data includes residual data and a set of cells including at least one primary cell determining a feature set of the pattern and at least one secondary cell determining a feature set of the pattern, the residual data includes primary subroutine call commands, the primary cell includes at least two secondary subroutine call commands, each of the subroutine call commands identifies a cell of the set of cells and a portion of the target at which the feature set determined by the cell is to be imaged; (b) storing each of the cells in a memory; and (c) generating the beam control data in response to the residual data and the contents of the memory, including by responding to each of at least two of the primary call commands by retrieving the primary cell from the memory and generating a portion of the beam control data which controls imaging of the feature set determined by the primary cell at the portion of the target identified by said each of at least two of the primary call commands, and responding to each of at least two of the secondary call commands by retrieving the secondary cell from the memory and generating a portion of the beam control data which controls imaging of the feature set determined by the secondary cell at the portion of the target identified by said each of at least two of the secondary call commands. 24. The method of claim 23 , wherein step (c) is performed such that the beam control data is in pixel format. claim 23 25. The method of claim 23 , also including the step of: claim 23 (d) in response to the beam control data, operating a beam system in a sequence of configurations which cause the at least one beam to image the pattern on the target. 26. The method of claim 25 , wherein the beam system executes a raster scan of the at least one beam relative to the target during step (d). claim 25 27. A method for generating beam control data for causing at least one beam to image a pattern on a target, said method comprising the steps of: (a) operating a graphics engine to receive a set of hierarchical image data having at least three levels of hierarchy and comprising a set of cells including at least one primary cell determining a feature set of the pattern and at least one secondary cell determining a feature set of the pattern, and residual data, and to store the primary cell and the secondary cell in a memory, wherein the residual data includes primary subroutine call commands, and the primary cell includes secondary call commands, wherein each of the subroutine call commands identifies one of the cells and a portion of the target at which the feature set determined by said one of the cells is to be imaged; and (b) operating the graphics engine to generate the beam control data in response to the residual data and the contents of the memory, including by responding to each of at least two of the primary call commands by retrieving the primary cell from the memory and generating a portion of the beam control data which controls imaging of the feature set determined by the primary cell at the portion of the target identified by said each of at least two of the primary call commands, and responding to each of at least two of the secondary call commands by retrieving the secondary cell from the memory and generating a portion of the beam control data which controls imaging of the feature set determined by the secondary cell at the portion of the target identified by said each of at least two of the secondary call commands. 28. The method of claim 27 , wherein step (b) is performed such that the beam control data is in pixel format. claim 27 29. The method of claim 27 , also including the step of: claim 27 (c) in response to the beam control data, operating a beam system in a sequence of configurations which cause the at least one beam to image the pattern on the target. 30. The method of claim 29 , wherein the beam system executes a raster scan of the at least one beam relative to the target during step (c). claim 29 31. A method for generating beam control data for causing at least one beam to image a pattern on a target, said method comprising the steps of: (a) operating a graphics engine to receive a set of hierarchical image data having at least three levels of hierarchy and comprising a set of cells including at least one primary cell and at least one secondary cell, and to generate and store in a memory a beam control data cell in response to each said primary cell and each said secondary cell of the hierarchical image data, wherein the hierarchical image data also includes residual data including subroutine call commands, and each of the subroutine call commands identifies one said beam control data cell and a portion of the target at which a feature set determined by said beam control data cell is to be imaged; and (b) operating the graphics engine to generate the beam control data in response to the residual data and the contents of the memory, including by responding to each of at least two of the subroutine call commands by retrieving one said beam control data cell from the memory. 32. The method of claim 31 , wherein step (b) is performed such that the beam control data is in pixel format. claim 31 33. The method of claim 32 , also including the step of: claim 32 (c) in response to the beam control data, operating a beam system in a sequence of configurations which cause the at least one beam to image the pattern on the target. 34. The method of claim 33 , wherein the beam system executes a raster scan of the at least one beam relative to the target during step (c). claim 33
	claims	1. A fuel rod auto-loading apparatus for a nuclear fuel assembly, the fuel rod auto-loading apparatus comprising:a fuel rod storage unit includinga plurality of stationary beams installed in a vertical direction, anda plurality of racks fixed to the plurality of stationary beams, wherein the racks are stacked one above the other and on which fuel rods are horizontally stored;a fuel rod loading unit configured to supply the fuel rods to the fuel rod storage unit;a fuel rod unloading unit configured to unload the fuel rods from the fuel rod storage unit, wherein the fuel rod loading unit and the fuel rod unloading unit are disposed in opposite sides of the fuel load storage unit with each other;a feeding unit configured to arrange the fuel rods transferred from the fuel rod unloading unit for loading the fuel rods into a fuel rod case;a fuel rod assembly lifter disposed collinearly in a longitudinal direction with the feeding unit, and configured to move upward or downward the fuel rod case separably placed on the fuel rod assembly lifter; anda controller controlling the fuel rod loading unit, the feeding unit, and the fuel rod unloading unit,wherein the fuel rod unloading unit, the fuel rod storage unit, the fuel rod unloading unit and the feeding unit are disposed in order. 2. The fuel rod auto-loading apparatus set forth in claim 1, wherein the fuel rod loading unit includes:a loading conveyer on which the fuel rod transferred from another process is placed;a tilting part handing over the fuel rod placed on the loading conveyer to the fuel rod storage unit; anda raising/lowering driver raising/lowering the loading conveyer. 3. The fuel rod auto-loading apparatus set forth in claim 2, wherein the tilting part includes a driving tilting part and a driven tilting part connected to the driving tilting part by a shaft so as to be cooperated. 4. The fuel rod auto-loading apparatus set forth in claim 3, wherein the driving tilting part includes:a fuel rod seat bar that is pivotably provided; anda seat bar driver connected to the fuel rod seat bar by a link in order to rotate the fuel rod seat bar within a predetermined angle range. 5. The fuel rod auto-loading apparatus set forth in claim 1, wherein the feeding unit includes:a feeding guide module that has a drive source so as to be able to move leftward/rightward relative to an upper portion of a main frame and a fuel rod seat member so as to allow the fuel rods to be placed in a row; anda feeding module that provides a driving force for loading the fuel rods placed on the feeding guide module into the fuel rod case. 6. The fuel rod auto-loading apparatus set forth in claim 5, further comprising a plurality of guide wings at the upper portion of the main frame in order to guide the fuel rods to the fuel rod seat member. 7. The fuel rod auto-loading apparatus set forth in claim 6, wherein the guide wings are provided in parallel in pairs, and a tilting member is provided between each pair of guide wings to transfer the fuel rods placed between the guide wings to the fuel rod seat member one by one. 8. The fuel rod auto-loading apparatus set forth in claim 7, wherein the tilting member includes:a tilting lever having a second protrusion spaced apart from a first protrusion protruding from the guide wing; anda driver vertically driving the tilting lever. 9. The fuel rod auto-loading apparatus set forth in claim 1, wherein the feeding unit further includes a bar code reader that reads a bar code marked on each fuel rod on a transfer path of the fuel rods loaded into the fuel rod case. 10. The fuel rod auto-loading apparatus set forth in claim 1, wherein each of the racks includes a plurality of rack bars fixed to the plurality of stationary beams in a transverse direction, such that the plurality of the rack bars are inclined from the fuel rod loading unit toward the fuel rod unloading unit at a predetermined angle.
046769358	claims	1. In a method for producing mixed-oxide fuel pellets of uranium dioxide and plutonium dioxide by milling a powder uranium dioxide and plutonium dioxide by milling a powder mixture of uranium dioxide with up to 50% by weight plutonium dioxide and by subsequent granulation, pressing and sintering, the improvement comprosing increasing the solubility of the pellet is nitric acid by effecting the milling of the powder mixture in an unsintered physical mixture of uranium dioxide and plutonium dioxide in the presence of a halogen-free organic milling aid consisting of propane diol which is expelled from the mixture by heating to a temperature up to said sintering, and wherein milling is conducted to produce a primary grain size of less than 2 micrometers and wherein said subsequent sintering is a single sintering operation of the unsintered physical mixture of uranium dioxide and plutonium dioxide.
	claims	1. A method of producing a cladding tube for nuclear fuel for a nuclear boiling water reactor, which method comprises the following steps:forming a tube comprising:an outer cylindrical component comprising a first composition comprising a zirconium-based alloy; andan inner cylindrical component metallurgically bonded to the outer component, wherein the inner component comprises a second composition comprising 0.1 to 0.4 percentage by weight Sn, 400 to 1500 ppm Fe, less than 600 ppm O and the rest Zr, except for impurities of a content that does not exceed that which is normally accepted in Zr or Zr-alloys for applications in nuclear reactors,wherein the first and second compositions materially differ from each other, and wherein the second composition has a lower recrystallization temperature than the first composition;rolling the tube; andfinally annealing the cladding tube at a temperature and a time that results in a first degree of recrystallization of the outer component and a second degree of recrystallization of the inner component, wherein the second degree is at least 97 percent and wherein the first degree is less than the second degree and higher than 50%. 2. A method according to claim 1, wherein the second degree of recrystallization is 100% and the first degree of recrystallization is between 50% and 96%. 3. A method according to claim 1, wherein the first composition comprises Zircaloy 2 or Zircaloy 4. 4. A method according to claim 1, wherein the inner component has a thickness such that it constitutes between 3% and 30% of the total thickness of the cladding tube. 5. A method according to claim 1, wherein the step of finally annealing is carried out at a temperature between 485° C. and 550° C. 6. A method according to claim 1, wherein the step of finally annealing is performed for 1 h to 6 h. 7. A cladding tube for nuclear fuel for a nuclear boiling water reactor, which cladding tube comprises:an outer cylindrical component comprising a first composition comprising a zirconium-based alloy and having a first recrystallization temperature, wherein the outer cylindrical component has a first degree of recrystallization higher than 50 percent; andan inner cylindrical component comprising a second composition comprising 0.1 to 0.4 percentage by weight Sn, 400 to 1500 ppm Fe, less than 600 ppm O and the rest Zr, except for impurities of a content that does not exceed that which is normally accepted in Zr or Zr-alloys for applications in nuclear reactors, and having a second recrystallization temperature lower than the first recrystallization temperature, wherein the inner cylindrical component has a second degree of recrystallization greater than the first degree of recrystallization and at least 97 percent, wherein the inner cylindrical component is metallurgically bonded to the outer component, and wherein the first and second compositions materially differ from each other. 8. A cladding tube according to claim 7, wherein the second degree of recrystallization is 100% and the first degree of recrystallization is between 50% and 96%. 9. A cladding tube according to claim 7, wherein the first composition comprises Zircaloy 2 or Zircaloy 4. 10. A cladding tube according to claim 7, wherein the inner component has a thickness such that it constitutes between 3% and 30% of the total thickness of the cladding tube. 11. A method according to claim 1, wherein the second degree of recrystallization is at least 97 percent and the first degree of recrystallization is between 70 percent and 90 percent. 12. A cladding tube according to claim 7, wherein the second degree of recrystallization is 100% and the first degree of recrystallization in the outer component is between 70 percent and 90 percent. 13. The method of claim 1, wherein the step of finally annealing is carried out at a temperature between 485° C. and 515° C. 14. The method of claim 1, wherein the second composition consists essentially of 0.1 to 0.4 percentage by weight Sn, 400 to 1500 ppm Fe, less than 600 ppm O and the rest Zr, except for impurities of a content that does not exceed that which is normally accepted in Zr or Zr-alloys for applications in nuclear reactors. 15. The cladding tube according to claim 7, wherein the second recrystallization temperature is between 485° C. and 550° C. 16. The cladding tube according to claim 15, wherein the second recrystallization temperature is between 485° C. and 515° C. 17. The method of claim 10, wherein the step of finally annealing is carried out at a temperature between 485° C. and 515° C. 18. The method of claim 1, wherein the first degree of recrystallization is between 60 percent and 90 percent. 19. The cladding tube of claim 7, wherein the first degree of recrystallization is between 60 percent and 90 percent. 20. A fuel assembly for a nuclear boiling water reactor, comprising:an enclosing tube; anda plurality of cladding tubes according to claim 7 filled with nuclear fuel, wherein said plurality of cladding tubes are arranged inside said enclosing tube.
	claims	1. A system for receiving and storing high level radioactive waste comprising:a concrete enclosure comprising walls, a roof and a floor that collectively form an internal space;the roof comprising an array of holes;an array of metal shells, each metal shell having a cavity for accommodating one or more containers holding high level radioactive waste, the array of metal shells arranged in a substantially vertical and spaced apart manner within the internal space of the enclosure, the array of the metal shells being co-axial with the array of holes of the roof so that containers holding high level radioactive waste can be lowered through the array of holes of the roof and into the cavities of the array of metal shells;the array of metal shells fastened to the floor and to the roof, the array of metal shells being load bearing columns for the roof;each of the metal shells comprising: (i) an expansion joint for accommodating thermal expansion and contraction of the metal shells; and (ii) one or more holes at a bottom portion of the metal shell that create a passageway between the internal space and the cavity of the metal shell; andthe concrete enclosure comprising one or more inlet ventilations ducts forming passageways from outside of the concrete enclosure to the internal space of the concrete enclosure. 2. The system of claim 1 wherein the internal space circumferentially surrounds each of the metal shells. 3. The system of claim 2 wherein each of the metal shells have a length; andwherein, for each metal shell, the internal space circumferentially surrounds a major portion of the length. 4. The system of claim 1 wherein, for each of the metal shells, the expansion joint is a flanged and flued section of the metal shell. 5. The system of claim 1 wherein, for each metal shell, the expansion, is a collar structure connected to the metal shell. 6. The system of claim 1 wherein, for a major portion of a length of each of the metal shells, a line of sight exists between outer surfaces of adjacent ones of the metal shells. 7. The system of claim 1 wherein the walls of the concrete enclosure are upstanding walls comprising an inner surface and an outer surface; wherein the one or more inlet ventilation ducts are located within upstanding walls; and wherein the one or more inlet ventilation ducts comprise a first inlet ventilation duct that forms a passageway from outside of the concrete enclosure to a top portion of the internal space, and a second inlet ventilation duct that forms a passageway from outside of the concrete enclosure to a bottom portion of the internal space of the concrete enclosure. 8. The system of claim 1 further comprising a gridwork of intersecting beams formed into and protruding from a bottom surface of the roof, the gridwork of beams surrounding the perimeter of each of the holes of the roof. 9. The system of claim 1 wherein, each of the walls, comprises two overlapping and inter-fitted wall structures. 10. The system of claim 1 further comprising:a plurality of containers holding high level radioactive waste positioned within the cavities of the array of metal shells;wherein the cavities of the array of the metal shells have a horizontal cross-section that accommodates no more than one of the containers; andwherein for each container positioned in the cavities, an annular gap exists between an outer surface of the container and the inner surface of the metal shell. 11. The system of claim 1 wherein each of the metal shells is fastened to the roof of the concrete enclosure so that a hermetically sealed interface exists between the metal shell and the roof. 12. The system of claim 1 further comprising:a container receiving area adjacent the concrete enclosure; anda crane system comprising a crane configured to translate between a position above the container receiving area and a position above the concrete enclosure. 13. The system of claim 12 wherein the crane system further comprises: a frame structure extending from the concrete enclosure and into the container receiving area;rails extending along, the roof of the concrete enclosure and the frame structure; andwherein the crane is operably coupled atop the rails. 14. The system of claim 1 wherein, for each metal shell, the expansion joint is located above the one or more holes of the metal shell and below the roof of the concrete enclosure. 15. The system of claim 1 further comprising;for each of the metal shells, a first tubular section that forms the cavity, a flange surrounding a top of the first tubular section, and a second tubular section extending upward from an outer edge of the flange;wherein each of the holes of the roof is formed by a stepped surface having a first riser surface, a tread surface, and a second riser surface; andwherein the array of metal shells extend through the array of holes of the roof so that the flanges of the metal shells rest on the tread surfaces of the array of holes of the roof and the second tubular section protrudes from a top surface of the roof. 16. The system of claim 15 further comprising:for each of the metal shells, a plurality of circumferentially spaced brackets atop the flange; andfor each of the metal shells, a lid positioned atop the plurality of brackets, the plurality of brackets supporting the lid in a spaced relationship from both the flange and the second tubular section so as to create passageways from the cavity to outside of the concrete enclosure. 17. A system for receiving and storing high level radioactive waste comprising:an enclosure comprising walls having inlet ventilation ducts, a roof comprising an array of holes, and a floor;an array of metal shells located in an internal space of the enclosure, the array of metal shells being co-axial with the array of holes in the roof so that containers holding high level radioactive waste can be lowered through the array of holes of the roof and into the array of metal shells;the array of metal shells being load bearing columns for the roof; andeach of the metal shells comprising one or more holes at a bottom portion of the metal shell. 18. The system of claim 17 wherein each of the metal shells further comprises an expansion joint for accommodating thermal expansion and/or contraction of the metal shells. 19. The system of claim 17 wherein each of the metal shells have a length;wherein the internal space is an uninterrupted volume that circumferentially surrounds each of the metal shells for a major portion of the length; and wherein a line of sight exists between outer surfaces of adjacent ones of the metal shells for the major portions of the lengths. 20. A system for receiving and storing high level radioactive waste comprising:an enclosure comprising; walls, a roof comprising an array of holes, and a floor;an array of load bearing columns in an internal space of the enclosure that support the roof, each of the load hearing columns comprising:a metal shell forming a cavity that is aligned with one of the holes of the array of holes so that a container holding high level radioactive waste can be lowered through the one of the holes of the roof into the cavity; andone or more holes at a bottom portion of the metal shell;the enclosure comprising one or more inlet ventilations ducts forming passageways from outside of the enclosure to the internal space;for each cavity, one or more outlet passageways extending from a top of the cavity to outside of the concrete enclosure; andfor each hole of the roof, a lid covering, the hole of the roof that prevents radiation from escaping via the hole of the roof. 21. The system of claim 1 further comprising, for each cavity, one or more outlet passageways extending from a top of the cavity to outside of the concrete enclosure. 22. The system of claim 21 further comprising, for each hole of the roof, a lid covering the hole of the roof that prevents radiation shine from the hole of the roof.
	abstract	An apparatus for x-raying a semiconductor device which includes semiconductor material and conductive material, the apparatus including a source of x-rays, a filter for receiving x-rays from the source of x-rays and allowing transmission of x-rays to the device, the filter having an atomic number greater than the atomic number of the conductive material of the device, and an x-ray imager for receiving x-rays from the device.
	claims	1. A casing internal part suitable for the dry intermediate storage of at least one irradiated fuel element, characterized in that it has a tiered structure which comprises at least two superposed modules, each of said modules, made of a material that is a good conductor of heat, with compartment(s) for accommodating said fuel element(s) in the central part and with at least one hollowed-out peripheral heat sink, being arranged on a support plate that is perforated to allow said fuel element(s) through; the perforated support plates of said structure being positioned and assembled by means of a retaining system, with a clearance left between the top of the module of one tier and the perforated support plate of the tier immediately above. 2. The casing internal part as claimed in claim 1, characterized in that said at least one compartment and said at least one hollowed-out pheriphal heat sink of each module are based on extruded elements. 3. The casing internal part as claimed in claim 2, characterized in that each module of its tiered structure comprises at least one prefabricated compartment. 4. The casing internal part as claimed in claim 1, characterized in that said at least one hollowed-out peripheral heat sink is an extruded element with or without fin(s) and with radiating surface(s). 5. The casing internal part as claimed in claim 1, characterized in that each module of its tiered structure comprises 1, 3, 4, 7, 12 or 15 compartments. 6. The casing internal part as claimed in claim 1, characterized in that said modules are made of aluminum; and/or, advantageously and, said retaining system and said perforated support plates are made of a material with high mechanical strength and are advantageously made of stainless steel. 7. The casing internal part as claimed in claim 1, characterized in that said retaining system comprises through bolts for, on the one hand, positioning and assembling the support plates and, on the other hand, transferring load from said casing internal part to the casing in which it is intended to be housed. 8. The casing internal part as claimed in claim 7, characterized in that said through bolts position each module on its support plate. 9. The casing internal part as claimed in claim 1, characterized in that its tiered structure comprises 2, 3, 4, 5 or 6 superposed modules. 10. The casing internal part as claimed in claim 1, characterized in that said at least one compartment has a square, hexagonal or circular cross section. 11. A casing suited to the dry intermediate storage of at least one irradiated fuel element, characterized in that it contains a casing internal part as claimed in claim 1 and comprises a bottom capable of withstanding the mass of said casing internal part and that of said at least one fuel element that is going to rest against it. 12. The casing as claimed in claim 11, characterized in that said casing and its casing internal part, which advantageously adopt the shape of right cylinders, have cross sections of roughly the same dimension(s). 13. A method for the dry intermediate storage of at least one irradiated fuel element comprising:the stable conditioning of said at least one irradiated fuel element in a casing within which the heat released by said at least one element is dissipated with no adverse effect on the structure of said at least one element,the intermediate storage of said sealed casing in a vertical pit cooled by the circulation of air,characterized in that said casing is a casing as claimed in claim 1. 14. The method as claimed in claim 13, characterized in that said casing contains several fuel elements. 15. The method as claimed in claim 13, characterized in that said at least one irradiated fuel is an element the power of which is 5 kW or lower.
048204730	abstract	The present invention is concerned with a method of reducing radioactivity in a nuclear plant by preliminarily forming oxide films on the surfaces of metallic structural members to be in contact with high-temperature and high-pressure reactor water containing radioactive substances before said metallic members are exposed to said reactor water. The method is characterized by the steps of subjecting said structural members to a first-step oxidation treatment of heating said structural members in an environment of a high temperature, and further subjecting the thus treated structural members to a second step oxidation treatment of heating said treated structural members in an environment having a higher oxidizing capacity than that of said environment in said first-step oxidation treatment to form a denser oxide film than an oxide film obtained in said first step oxidation treatment. According to the present invention, radioactivity in the nuclear plant can be reduced remarkably.
	claims	1. A process for the treatment of a solvent which has been used in nuclear fuel reprocessing or uranium ore purification and which comprises an organophosphate ester and a hydrocarbon diluent, the process comprising distilling the solvent under reduced pressure to remove substantially all the diluent and a major proportion of the organophosphate ester and wherein a residual material including organophosphate is also formed, converting organophosphate present in the residual material to inorganic phosphate and encapsulating the inorganic phosphate. 2. A process according to claim 1 in which the distillate, containing substantially all of the diluent and a major proportion of the organophosphate ester, is returned to the reprocessing or purification process. claim 1 3. A process according to claim 1 in which said encapsulating step is vitrification. claim 1 4. A process according to claim 1 in which the organophosphate is converted to inorganic phosphate by a temperature process or a chemical oxidation process, wherein the temperature process comprises mixing the residual material with a metal salt hydroxide in aqueous solution to form a mixture and then treating the mixture to produce a metal phosphate ash. claim 1 5. A process according to claim 4 in which the chemical oxidation process includes using a metallic catalyst and hydrogen peroxide. claim 4 6. A process according to claim 5 in which the chemical oxidation process is carried out at a temperature between ambient and boiling point in an aqueous medium. claim 5 7. A process according to claim 1 in which the hydrocarbon diluent is odourless kerosene or a dodecane. claim 1
	summary
	abstract	A calibrated radioactive source comprises a container and a material labelled by at least one radionuclide. The labelled material is contained in the container and the container is made of a material that is transparent to the radiation emitted by the at least one radionuclide. The source is characterized in that the labelled material is a self-hardening polymer that is chemically inert relative to the material used for the container and in that the container is a flexible sheath. The calibrated source is placed into a hole of a brick of a tissue-equivalent phantom. An assembly is formed by such a calibrated source and a brick of tissue-equivalent phantom.
052271293	summary	BACKGROUND OF THE INVENTION This invention relates generally to fuel rods employed in nuclear reactors. More particularly, the present invention relates to fuel rods having a zirconium-alloy cladding tube which contains fuel pellets. Fuel rods having outer cladding tubes are mounted in support grids of the reactor fuel assembly. Because of the harsh environment of the fuel assembly where the surrounding water temperature is typically 400.degree. C. and the water has a relatively high pressure, the cladding tube is susceptible to wear and corrosion. At the lower portions of the reactor assembly, the cladding tubes are exposed to debris fretting. In addition, there are severe wear forces at the location of the grid support. A number of advancements have been introduced in some industrial applications to improve the ability of metallurgical thin film to combat wear, to resist chemical corrosion, to protect substrates from hostile environments and to resist erosion. For example, ion-assisted vacuum deposition techniques such as cathodic arc plasma deposition (CAPD) have been employed for depositing thin films on substrates to be protected. CAPD processes have achieved superior film bonding and higher densities than more conventional ion plated films. The conventional CAPD system includes a vacuum chamber, a cathode, an arc power supply, means of igniting an arc on the cathode surface, an anode and a substrate bias power supply. A vacuum arc is employed to evaporate the source material which functions as the cathode in the arc circuit. A voltage in the range of 15 to 50 volts is typically employed to sustain the arc. The voltage level is dependent upon the cathode material. Arcing is initiated by applying a high voltage pulse to an electrode near the cathode and/or by mechanical ignition. Evaporation occurs due to high velocity arc spots traversing across the cathode surface at velocities as great as 100 m/second. The arc spots carry high current densities and are sustained by the plasma that is generated by the arc. The high current density results in flash evaporation of the source material. The resulting vapor consist of electrons, ions, neutral vapor atoms and microdroplets. The electrons are accelerated toward a cloud of positive ions. Emissions from the cathode spots remain relatively constant over a wide range of arc currents as the cathode spots split into a multiplicity of spots. The average current carried per spot depends on the nature of the cathode material. SUMMARY OF THE INVENTION Briefly stated, the invention in a preferred form is a cladding tube for a nuclear fuel rod which has a thin film of zirconium nitride coating the outside surface. The zirconium nitride coat may be applied on a portion of the tube which will be located below or in the vicinity of a particular fuel assembly support grid, or on substantially the entire outside surface of the tube. A film having a thickness of approximately 5 microns is effective in resisting corrosion and wear of the cladding tube, which usually has a zirconium-alloy composition. The thin film of zirconium nitride is applied to the cladding tube by reactively depositing zirconium nitride on the surface of the cladding tube by a cathodic arc plasma deposition process. The cladding tube is heated to a temperature in the range of approximately 300.degree. to 400.degree. C. in the presence of nitrogen in a vacuum chamber. An object of the invention is to provide a new and improved corrosion resistant coating for a cladding tube employed in a nuclear reactor. Another object of the invention is to provide a new and improved coating which may be applied to a zirconium-alloy cladding tube in an efficient and cost effective manner. A further object of the invention is to provide a new and improved coating for a cladding tube which has enhanced resistance to debris fretting and corrosion and has an outer diameter which does not sufficiently increase the fuel assembly pressure drop. Other objects and advantages of the invention will become apparent from the drawings and the specification.
	claims	1. A device for selecting one of several triggering apparatuses which are simultaneously connectable to said device, said triggering apparatuses being arranged for producing each a triggering signal to enable/disable one or more components of a radiation treatment apparatus, said triggering signals depending on detected parameters, the device comprising:means for receiving triggering signals from said several triggering apparatuses, when all of said apparatuses are connected to the device,input means for selecting one of said triggering apparatuses,means for generating a universal triggering signal for said one or more components on the basis of said received triggering signal from said selected triggering apparatus,means for sending the universal triggering signal to said one or more components. 2. The device according to claim 1, wherein said input means for selecting one of said triggering apparatuses comprises a control panel having a switch to manually select the triggering apparatus. 3. The device according to claim 2, wherein said switch is a turnable knob, arranged to select one of said triggering apparatuses or a manual triggering method, or a state wherein no triggering of the beam is performed. 4. The device according to claim 1, wherein the device comprises means to select a manual triggering method, and a switch to manually enable/disable a component. 5. The device according to claim 1, wherein said input means for selecting one of said triggering apparatuses comprises means for receiving the selection of the triggering apparatus from an external system and for automatically selecting the triggering apparatus according to the received selection. 6. The device according to claim 1, wherein said input means, said means for generating and said means for sending are adapted for selecting, generating and sending a permanently enabled triggering signal. 7. The device according to any claim 1, wherein the device further comprises selection means for selecting a component from said one or more components. 8. The device according to claim 7, wherein said selection means for selecting a component comprises a control panel having a switch to select the component. 9. The device according to claim 7, wherein said selection means for selecting a component from said one or more components comprises means for receiving from an external system the selection of the component and for automatically selecting the component according to the received selection. 10. The device according to claim 1, the device further comprising:means for sending signals to the said several triggering apparatusesmeans for receiving signals from said one or more components. 11. The device according to claim 1, comprising for each triggering apparatus connectable to the device, a pair of switches being arranged in series, the first of said pair of switches being arranged to close when the corresponding triggering apparatus is selected, the second of said pair of switches being arranged to close or open in accordance with a triggering signal received from the corresponding triggering apparatus, the device further being arranged so that said universal triggering signal is generated when both switches are closed. 12. A method for treating a patient by using a device according to claim 1, said apparatus comprising a beam delivery system, the method comprising the steps of:selecting one of said triggering apparatuses connected to the device;receiving a triggering signal from the selected triggering apparatus;generating a gating signal for enabling/disabling said beam delivery system in accordance with said triggering signal; andirradiating said patient with said beam delivery system in accordance with said gating signal. 13. The method according to claim 12, wherein said radiation treatment apparatus comprises several beam delivery systems, and wherein the method comprises the steps of:selecting one of said triggering apparatuses connected to the device;receiving a triggering signal from the selected triggering apparatus;selecting a beam delivery system;generating a gating signal for enabling/disabling said beam delivery system in accordance with said triggering signal; andirradiating said patient with said beam delivery system in accordance with said gating signal. 14. The method according to claim 12, wherein said radiation treatment apparatus further comprises a patient positioning verification system and wherein said steps are preceded by the steps of:selecting one of said triggering apparatuses connected to the device;selecting said patient positioning verification system;receiving a triggering signal from the selected triggering apparatus;generating a gating signal for enabling/disabling said patient positioning verification system; andperforming said patient positioning verification in accordance with said gating signal. 15. The method according to claim 12, wherein said radiation treatment apparatus further comprises a patient positioning verification system and wherein said steps and the steps of irradiating said patient and of performing said patient positioning verification are performed simultaneously. 16. A device for selecting one of a plurality of triggering apparatuses which are simultaneously connected to the device, the triggering apparatuses configured to produce triggering signals to enable/disable one or more components of a radiation treatment apparatus, the triggering signals depending on detected parameters, the device comprising:a controller which receives the triggering signals from the plurality of triggering apparatuses,an input device which receives an input for selecting one of the plurality of triggering apparatuses,a generating device which generates a universal triggering signal for the one or more components on the basis of the received triggering signal from the selected triggering apparatus, andan output device which sends the universal triggering signal to the one or more components. 17. A method for treating a patient by using a device, the device being coupled to two or more beam triggering apparatuses and to a radiation treatment apparatus comprising a beam delivery system, the method comprising the steps of:selecting one of the two or more triggering apparatuses connected to the device;receiving a triggering signal from the selected triggering apparatus;generating a signal for enabling/disabling the beam delivery system in response to receiving the triggering signal; andirradiating a patient with the beam delivery system in accordance with said gating signal.
	claims	1. A method for identifying one or more malfunctions from a plurality of malfunctions that can occur in a locomotive air compressing system comprising a plurality of components, with at least some of said plurality of malfunctions being correctable onboard a locomotive, said at least some malfunctions constituting onboard serviceable malfunctions, while a remaining of said plurality of malfunctions are correctable with the air compressing system being uninstalled and serviced at an off-board servicing site, said remaining malfunctions constituting off-board serviceable malfunctions, said method comprising:performing a test sequence configured to isolate components of the air compressing system from one another to identify a component subject to a malfunction;applying a pressurizing stimuli to the isolated components;monitoring a response indicative of a malfunction type, the malfunction type being associated with one of the following: a corrective action to be performed onboard the locomotive, and a corrective action to be performed off-board the locomotive; andperforming the corrective action appropriate for the malfunction type. 2. The method of claim 1 further comprising performing a crankcase inspection prior to performing said test sequence, and determining whether or not predefined criteria for passing said crankcase inspection is met, and, if said crankcase inspection meets the predefined passing criteria, proceeding to perform the test sequence. 3. The method of claim 1 wherein said air compressor system comprises a high-pressure stage and a low-pressure stage and performing the test sequence comprises isolating the low-pressure stage from the high-pressure stage to identify a malfunction likely to correspond to a respective one of said stages. 4. The method of claim 3 wherein said low-pressure stage comprises at least one low-pressure cylinder and at least one intercooler coupled to the low-pressure cylinder and wherein the applying of the pressurizing stimuli comprises pressurizing through an inlet port said least one low-pressure cylinder and said at least one intercooler. 5. The method of claim 4 wherein the monitoring of a response indicative of a malfunction type comprises monitoring a depressurization rate of the pressurized said at least one low-pressure cylinder and said at least one intercooler. 6. The method of claim 5 further comprising comparing the monitored depressurization rate relative to a predefined depressurization rate limit, and identifying a likely malfunction type based on the results of said comparison. 7. The method of claim 4 wherein the monitoring of a response indicative of a malfunction type comprises visually monitoring said at least one intercooler to detect an air leak in the intercooler, and, in the event said intercooler air leak is identified, replacing said intercooler onboard the locomotive. 8. The method of claim 4 wherein the monitoring of a response indicative of a malfunction type comprises monitoring air flow through an oil-filling port in a crankcase, and wherein a sensing of said air flow is indicative of a leak through a wall of said at least one low-pressure cylinder, and, in the event said air flow is sensed, performing a compressor removal from the locomotive for performing a compressor overhaul at the specialized servicing site. 9. The method of claim 3 wherein said high-pressure stage comprises a high-pressure cylinder and an aftercooler coupled to the high-pressure cylinder and wherein the applying of a pressurizing stimuli comprises pressurizing through an outlet port said high-pressure cylinder and said aftercooler. 10. The method of claim 9 wherein the monitoring of a response indicative of a malfunction type comprises monitoring a depressurization rate of the pressurized high-pressure cylinder and aftercooler. 11. The method of claim 10 further comprising comparing the monitored depressurization rate relative to a predefined depressurization rate limit, and identifying a likely malfunction type based on the results of said comparison. 12. The method of claim 9 wherein the monitoring of a response indicative of a malfunction type comprises visually monitoring said aftercooler to detect an air leak in the aftercooler, and, in the event said aftercooler air leak is detected, replacing said aftercooler onboard the locomotive. 13. The method of claim 9 wherein the monitoring of a response indicative of a malfunction type comprises monitoring air flow through an oil-filling port in a crankcase, a sensing of said air flow being indicative of a leak through a high-pressure cylinder wall, and, in the event said air flow is sensed, performing a compressor removal from the locomotive for performing a compressor overhaul at the specialized servicing site. 14. The method of claim 9 wherein the monitoring of a response indicative of a malfunction type comprises sensing air flow through an intake port of the high pressure cylinder, a sensing of said air flow being indicative of a leak through a high-pressure cylinder wall being indicative of at least one malfunctioning valve in the high pressure cylinder, and, in the event said air flow is sensed, performing a cylinder head replacement onboard the locomotive for the high pressure cylinder. 15. The method of claim 1 wherein the applying of pressurizing stimuli comprises pressurizing a crankcase and the monitoring of a response indicative of a malfunction type comprises monitoring a depressurization rate of the pressurized crankcase, said crankcase pressurizing and monitoring avoiding removal of a compressor motor for gaining visual access to crankcase seals wherein leakage is likely to develop. 16. The method of claim 1 wherein the applying of pressuring stimuli comprises energizing a compressor motor at a predefined RPM and monitoring a volume of pressurized air actually delivered by said compressor system over a period of time, and comparing the volume of pressurized air delivered by the compressor system relative to a predefined range indicative of whether or not said air compressor system meets a specified air-compressing capability. 17. The method of claim 16 further comprising monitoring an intercooler pressure and comparing said intercooler pressure relative to predefined pressure values to determine occurrence of a likely malfunction in one of the following: a cylinder head of a high-pressure stage and cylinder heads of a low-pressure stage of said compressor system. 18. A method for identifying one or more malfunctions from a plurality of malfunctions that can occur in a vehicular air compressing system comprising a plurality of components, with at least some of said plurality of malfunctions being correctable onboard a self-propelled vehicle, said at least some malfunctions constituting onboard serviceable malfunctions, while a remaining of said plurality of malfunctions are correctable with the air compressing system being uninstalled and serviced at an off-board servicing site, said remaining malfunctions constituting off-board serviceable malfunctions, said method comprising:performing a test sequence configured to isolate components of the air compressing system from one another to identify a component subject to a malfunction;applying a pressurizing stimuli to the isolated components;monitoring a response indicative of a malfunction type, the malfunction type being associated with one of the following: a corrective action to be performed onboard the vehicle, and a corrective action to be performed off-board the vehicle; andperforming the corrective action appropriate for the malfunction type.
	abstract	A device operable to accelerate a particle beam to an energy for irradiating a target volume. The device includes a particle accelerator operable in a first working phase in which particles of the particle beam are accelerated to the energy and a second working phase in which the particles of the particle beam are provided and extracted for irradiating the target volume. The device further includes a control device operable to interrupt an irradiation of the target volume if the target volume assumes a predetermined state. The control device is also operable to control the particle accelerator as a function of a comparison between a residual particle number stored in the accelerator and a reference value.
040452840	summary	BACKGROUND OF THE INVENTION In a nuclear reactor, there is a danger that the nuclear fuel might get hot enough to melt and escape its containment, which would cause the release of large amounts of deadly radiation into the environment. Ordinarily reactor's are provided with water cooling ducts and other apparatus for maintaining the temperature of the fuel of the reactor, however, in the event of a failure of the cooling system, the reactor fuel will elevate the temperature to a point where the fuel will become melted into a flowing mass. This fluid flowing mass, will in turn, tend to increase its own temperature while it continues as a mass. Such a mass of molten fuel is extremely radioactive and accordingly, causes its own temperature elevation to a point where primary and secondary containments will be breached and will thereby expose the radioactive material to the environment. SUMMARY OF THE INVENTION It is, therefore, an object of this invention to provide a means for preventing escape of radioactive material into the environment from a nuclear reactor in which a gross core melt down has occurred. In this regard, this invention provides a reactor shield having a fuel containment chamber adapted to contain an atomic reactor and the reactor fuel therein. However, in the event of a gross core melt down of the reactor, the reactor shield is provided with an exhaust passage means whereby the melt down fuel and general radioactive mass produced thereby, will pass from the reactor shield and into a deactivating containment base provided by this invention. The deactivating containment base is provided with a dispersal means for receiving the radioactive mass of the melt down as it flows from the reactor and to distribute the flowing mass to separate the fuel into a plurality of smaller masses, which are preferably subcritical, whereby nuclear interaction, heat production, and dangerous isotope production are thereby reduced, the containment base acting as shielding between the masses and also functioning as a heat sink. Another embodiment of the invention further provides fuel fluxing material in the base dispersion means whereby the dangerously temperature elevated fuel mass, which is dispersed within the base, will be further cooled by the action of dissolving fluxing material with the fuel mass being dispersed within the base and thereby further provide for means for cooling the fuel. The fluxing and control material are maintained in a position apart from the core to eliminate uncertainties involving prolonged exposure of such materials to the core.
047449390	claims	1. A method for correcting for burn-in in a fissionable neutron dosimeter employing a fissionable isotope (Z, A), comprising the steps of: (a) forming two solid state track recorders with fission deposits of the same fissionable material as the fissionable neutron disometer; (b) exposing a first of the two solid state track recorders and the fissionable neutron dosimeter to a first neutron fluence, at least effectively with the same neutron flux-time history with respect to the location of one of them in the first neutron fluence, whereafter the fissionable neutron dosimeter indicates a total number of fissions F.sub.T which is to be corrected for the burn-in; (c) irradiating the two solid state track recorders with a second neutron fluence; (d) determining the amount of burn-in P.sub.Z', A' of a higher order isotope (Z', A'), wherein A'>A, in the fission deposit of the first solid state track recorder from the difference between the absolute numbers of fissions per unit volume of the fission deposits in the two solid state track recorders; (e) determining the number of fissions F.sub.Z', A' of the higher order isotope (Z', A') in the fissionable deposit of the first solid state track recorder during the exposure to the first neutron fluence; and (f) using P.sub.Z', A' and F.sub.Z', A' to correct the total number of fissions F.sub.T indicated by the fissionable neutron dosimeter, to provide a value corresponding to the fission rate of the fissionable isotope that is corrected for the fissions due to the burn-in. 2. The method as recited in claim 1, wherein the first neutron fluence is time independent or a separable function of time and neutron energy. 3. The method as recited in claim 2, wherein the step (e) of determining the number of fissions F.sub.Z', A' of the higher order isotope (Z', A') further comprises the substep of exposing a further fission neutron dosimeter prepared for the higher order isotope (Z', A') to at least effectively the same neutron flux-time history at the same location in the first neutron fluence. 4. The method as recited in claim 2, wherein the fission neutron dosimeter for which the burn-in is to be corrected is selected from the group consisting of a radiometric fission dosimeter and a solid state track recorder fission dosimeter. 5. The method as recited in claim 3, wherein the further fission neutron dosimeter is selected from the group consisting of a radiometric fission dosimeter and a solid state track recorder fission dosimeter. 6. The method as recited in claim 2, wherein the step (b) further comprises the substep of exposing the first solid state track recorder fission deposit and the fissionable neutron dosimeter to be corrected for the burn-in at separate times but for corresponding periods and neutron fluxes at the same location. 7. The method of claim 1, further comprising the step of determining the value of the first neutron fluence based on the total number of fissions F.sub.T as corrected for the burn-in.
	claims	1. A deposition substrate comprising a support and a reflective layer disposed on the support,the reflective layer including light-scattering particles and a binder resin,the light-scattering particles present in a region extending in a thickness of from 0 to 0.5 μm from the surface of the reflective layer opposite to the surface in contact with the support toward the surface in contact with the support having an area average particle diameter of not more than 0.5 μm. 2. The deposition substrate according to claim 1, wherein the light-scattering particles include at least one selected from alumina, yttrium oxide, zirconium oxide, titanium dioxide, barium sulfate, silica, zinc oxide, calcium carbonate, glasses and resins. 3. The deposition substrate according to claim 1, wherein the light-scattering particles include at least one type of particles selected from hollow particles having a hollow portion within the particle, and porous particles. 4. The deposition substrate according to claim 1, wherein the light-scattering particles include at least titanium dioxide. 5. A scintillator panel comprising the deposition substrate described in claim 1, and a scintillator layer formed on the reflective layer of the deposition substrate by deposition, the scintillator layer including cesium iodide and at least one activator selected from at least thallium compounds, sodium compounds and indium compounds, the scintillator layer having a columnar crystal structure. 6. The scintillator panel according to claim 5, wherein the entire surface of the scintillator layer and a portion of the reflective layer are covered with a continuous protective film. 7. The scintillator panel according to claim 6, wherein the protective film is a protective film formed by a gas-phase method and includes at least one material selected from polyparaxylylene, polyurea and silicon dioxide (SiO2).
	abstract	An intermediate end plug assembly for a segmented fuel rod can stably support the fuel rod to the end of its cycle even if an interval between the fuel rods becomes narrow due to application of a dual-cooled fuel rod, and reduce excess vibration induced by flows of interior and exterior channels of the dual-cooled fuel rod for obtaining high burnup and output. To this end, the fuel rod has a segmented structure so as to make its length short. A lower intermediate end plug includes at least one channel hole, through which a coolant flows into an internal channel of the fuel rod, so that a possibility of causing departure from nuclear boiling ratio (DNBR) of the dual-cooled fuel rod is reduced.
	claims	1. A method of manufacturing a reactor internal structure, comprising the steps of:applying a titanium-zirconium compound solution to at least a part of a surface of the reactor internal structure constituting a boiling water reactor, the part being exposed to a reactor water; andforming a coating of amorphous zirconium titanate by heat-treating the exposed part of the surface of the reactor internal structure coated with the solution, so that the surface of the reactor internal structure has a negative surface potential. 2. The method of manufacturing a reactor internal structure according to claim 1, wherein the titanium-zirconium compound solution has a one-to-one atomic ratio of titanium to zirconium. 3. The method of manufacturing a reactor internal structure according to claim 1, wherein the heat treatment is performed at 80° C. to 600° C. 4. The method of manufacturing a reactor internal structure according to claim 1, wherein the coating has a thickness of 0.01 μm to 10 μm.
	description	The present invention provides a wafer holder assembly that is well-suited for SIMOX wafer processing, which includes the use of relatively high ion beam energies and temperatures. In general, the wafer holder assembly has a structure that maintains its integrity and reduces the likelihood of wafer contamination during extreme conditions associated with SIMOX wafer processing. The wafer holder assembly can be formed from electrically conductive materials to provide an electrical path from the wafer to ground for preventing electrical charging of the wafer, and possible arcing, during the ion implantation process. FIGS. 1-2 show a wafer holder assembly 100 in accordance with the present invention. The assembly includes first and second main structural rail members 102,104 that are substantially parallel to each other and spaced apart at a predetermined distance. In the exemplary embodiment shown, the main structural members 102,104 are generally C-shaped. A first wafer-holding arm 106 is rotatably secured to a distal end 108 of the holder assembly and a second wafer-holding arm 110 is pivotably secured to the assembly at a generally proximal region 112 of the assembly. The first arm 106 includes a transverse member 114 having first and second portions 116,118 each of which terminates in a respective distal end 120,122. Wafer-contacting pins 124,126 are secured to the distal ends 120,122 of the first and second arm portions. The first arm 106 is rotatable about a first axis 128 that is generally parallel to the first and second main structural members 102,104. By allowing the first arm 106 to rotate about the first axis 128, the first and second arm portions apply substantially equal pressure to the wafer edge via the spaced apart wafer-contacting pins 124,126. The second arm 110 is pivotable about a second axis 130 that is generally perpendicular to the main structural members 102,104 to facilitate loading and unloading of the wafers. A wafer-contacting pin 132 is affixed to the distal end 134 of the second arm to provide, in combination with the pins 124,126 coupled to the first arm, three spaced apart contact points to securely hold the wafer in place. Typically, placement of the pins about the circumference of the wafer is limited by a notch or xe2x80x9csignificant flatxe2x80x9d in the wafer that is used for orientating the wafer on the holder assembly. Some processing techniques include rotating the wafer a quarter turn, for example, one or more times during the implantation process to ensure uniform doping levels. The wafer holder assembly can further include a series of retaining members for securing the components of the assembly together without the need for conventional fasteners and/or adhesives. It is understood that adhesives can vaporize or outgas during the ion implantation process and contaminate the wafer. Similarly, conventional fasteners, such as exposed metal screws, nuts, bolts, and rivets can also contaminate the wafer. In addition, such devices may have incompatible thermal coefficients of expansion making the assembly prone to catastrophic failure. In one embodiment, the assembly includes a distal retaining member 136 coupling the first arm 106 to the assembly and an intermediate retaining member 138 affixed to a bottom of the assembly to maintain the spacing of the first and second main structural members 102,104 in a middle region 140 of the assembly. The assembly can further include a proximal retaining member 142 securing the structural members in position at the proximal region 112 of the assembly. FIGS. 3-7 (shown without the wafer-contacting pins), in combination with FIGS. 1 and 2, show further details of the wafer holder assembly structure. The first arm 106 includes a support member 144 extending perpendicularly from the transverse member 114 (FIGS. 3-4). The support member 144 includes an intermediate region 146 and an arcuate coupling member 148. A bearing number 150 extends through a longitudinal bore 152 in the intermediate region 146 of the support member 144 (FIGS. 3-4). A first cross member 154 is matable with the distal ends 156,158 of the main structural members 102,104 and a second cross member 160 is matable to the main structural members at a predetermined distance from the first cross member 154 (FIGS. 5-6). The first and second cross members 154,160 are adapted for mating with opposite edges of the main structural members 102,104. It is understood that notches can be formed in the various components to receive mating components. Each of the first and second cross members 154,160 includes a respective bore 162,164 for receiving an end of the bearing member 150. (FIG. 7). In one embodiment, the bearing member is a rod having each end seated within respective sleeve members 166,168 disposed within an aperture in the cross members 154,160. The sleeve members 166,168 allow the first arm 106 to freely rotate while minimizing particle generation due to graphite on graphite contact during rotation of the first arm. In one embodiment, the sleeves are formed from a hard, insulative material, such as aluminum oxide (sapphire). FIG. 8, in combination with FIGS. 1 and 2, show further details of the distal retaining member 136 having a first end 170 with a first notch 172 for coupling to one of the main structural members 102 and a second notch 174 for engaging the coupling member 148 (FIG. 3) of the first arm. A second end 176 of the distal retaining member 136 is matable to the intermediate region 140 of the assembly. Indents 178 can be formed in the main structural members 102,104 to facilitate engagement of the second end 176 to the assembly (FIG. 1). FIGS. 9-10 show alternative embodiments of the distal retaining member in the form of a helical spring 136xe2x80x2 and a bellows 136xe2x80x3, respectively. It is understood that one of ordinary skill in the art can readily modify the geometry of the retaining members. In one embodiment, the distal retaining member 136 is under tension so as to apply a force having a direction indicated by arrow 180 (FIG. 5) on the coupling member 148 of the support member. The force applied by the distal retaining member 136 pressures a neck 182 (FIG. 3) of the support member against the second cross member 160. The applied force also pressures the first cross member 154, via the bearing member 150, against the main structural members 102,104 as the second cross member 160 functions as a fulcrum for the support member 144. However, the transverse portion 114, as well as the support member 144 of the first arm, freely rotate about the first axis 128, i.e., the bearing member 150, such that the pins 124,126 at the distal ends of the first arm portions 116,118 provide substantially equal pressure on the wafer. FIGS. 11 and 12 (bottom view), in combination with FIGS. 1 and 2, show further details of the second proximal region 112 of the wafer holder assembly 100. FIG. 11 is shown without the second main structural member 104 for clarity. First and second stop members 184 (FIG. 1), 186 extend from the main structural members 102,104. In an exemplary embodiment, the second arm 110 includes wing regions 188 (FIG. 1) that are biased against the ends of the stop members 184,186 by a bias member 190. In one embodiment, the bias member 190 is under compression so as to pressure the second arm 110 against the stop members 184,186, e.g., the wafer-hold position. The bias member 190 includes a U-shaped outer portion 192 having a first end 194 mated to the first structural member 102 and a second end 196 coupled to the second structural member 104 (FIG. 12). A spring portion 198 of the second bias member includes one end abutting the second arm member 110 and the other end extending from a bottom of the U-shaped outer member 192. The second arm 110 pivots at its bottom end about a second bearing member 200 disposed on the second axis 130, which is generally perpendicular to the main structural members 102,104. The second bearing member 200 extends through a bore in the second arm with each end of the bearing member being seated in a sleeve inserted within a respective main structural member 102,104. Rotation of the second arm 110 is limited by respective brace members 202,204 extending from the main structural members 102,104. FIG. 13 (bottom view), in combination with FIGS. 1 and 2, shows further details of the intermediate retaining member 138, which is mated to the main structural members 102,104 in the intermediate region 140 of the assembly. The intermediate retaining member 138 includes first and second opposing U-shaped outer members 206,208 with a spring member 210 extending therebetween. The first outer member 206 has first and second arms 212,214 for mating engagement with corresponding notched protrusions 216,218 formed on the bottom of the main structural members 102,104. Similarly, the second outer member 208 includes arms that are matable with notched protrusions 220,222. In one embodiment, the U-shaped outer members 206,208 are forced apart to facilitate mating to the protrusions. Upon proper positioning, the outer members 206,208 are released such that spring member 210 biases the outer members against the protrusions. The intermediate retaining member 138 is effective to maintain the spacing between the first and second main structural members 102,104 and enhance the overall mechanical strength of the assembly. FIG. 14 shows the proximal retaining member 142, which provides structural rigidity in the proximal region 112 of the wafer holder assembly. In one embodiment, the proximal retaining member 142 includes upper and lower members 224,226 coupled by a spring member 228. The spring member 228 can be engaged to the main structural members such that the spring member is under tension. The proximal retaining member 142 can include a protruding member 230 having a slot 232 formed therein. As shown in FIG. 15, the assembly 100 is matable with a rotatable hub assembly 250 to which a series of wafer holder assemblies can be secured. A shield 252 can be secured to the proximal region 112 of the assembly to protect exposed regions of the assembly from beam strike. The shield 252 prevents sputtering from the assembly components, as well as any metal devices used to affix the assembly to the hub 250, during the ion implantation process. In addition, the assembly components are not heated by direct exposure to the ion beam. In one embodiment, an edge of the shield 252 is captured in the slot 232 (FIG. 14) located in the proximal retaining member 142. It is understood that the shield 252 can have a variety of geometries that are effective to shield the assembly components from beam strike. In one embodiment, the shield 252 is substantially flat with an arcuate edge 254 proximate the second wafer-holding arm 110 to increase the shielded region of the assembly. It is further understood that the shield can be formed from various materials that are suitably rigid and are opaque to the ion beam. One exemplary material is silicon having properties that are similar to a silicon wafer. The wafer-contacting pins 124,126,132 coupled to ends of the wafer-holding arms are adapted for contacting and securing the wafer in the wafer holder assembly 100. In general, the pins should apply sufficient pressure to maintain the wafers in the holder assembly during the load and unload process in which the wafers are manipulated through a range of motion that can include a vertical orientation. However, undue pressure on the wafers should be avoided since damage to the wafer surface and/or edge can result in the formation of a slip line during the subsequent high temperature annealing process. In addition, the wafer-contacting pins should not electrically insulate the wafer from the assembly. Further, the pins should be formed from a material that minimizes contamination of the wafer. FIGS. 16A-B show a wafer-contacting pin 300 adapted for use with a wafer holder assembly in accordance with the present invention. The pin has a distal portion 302 having a geometry adapted for holding the edge of a wafer and a proximal portion 304 having a contour complementing a corresponding channel formed in the ends of the wafer arms 106,110 (FIG. 1). It is understood that a variety of shapes and surface features can be used to securely and releasably mate the pin 300 to the wafer-holding arms. The distal portion 302 of the pin includes a ridge 306 extending from an arcuate wafer-receiving groove 308 in the pin. A tapered surface 310 extends proximally from the groove 308. As shown in FIG. 17, the pin should contact the top 352 and bottom 354 of the wafer 350 to prevent movement and/or vibration of the wafer as the holder assembly is rotated during the implantation process. In addition, the tapered surface 310 provides a ramp on which the wafer edge may first contact and slide upon during the wafer load process until meeting the ridge 306. FIGS. 18-23 show a wafer-contacting pin 400 in accordance with the present invention having a more limited profile. The pin 400 includes a distal portion 402 for holding a wafer and a proximal portion 404 for coupling to the arm ends. The distal portion 402 of the pin is rounded to minimize the amount of pin material proximate the wafer edge for reducing the likelihood of electrical discharge from the wafer to the pin. In addition, the pin geometry is optimized to maximize the distance between the wafer edge and the pin except at the wafer/pin contact interface. Further, the wafer-contacting region of the pin 400 should be smooth to minimize the electric field generated by a potential difference between the wafer and the pin. The pin should also minimize the wafer/pin contact area. The distal portion 402 of the pin includes a wafer-receiving groove or neck 406 disposed between a wedge-shaped upper region 408 and a tapered surface 410. The neck 406 can be arcuate to minimize the contact area between the wafer edge and the pin. The upper region 408, including the neck 406, can taper to a point or edge 412 for reducing the amount of pin material near the wafer edge to inhibit electrical arcing between the wafer and the pin. It is understood that the term wedge-shaped should be construed broadly to include a variety of geometries for the pin upper region. In general, the wedge-shaped upper region broadens from a point nearest a center of a wafer held in the assembly. Exemplary geometries include triangular, arcuate, and polygonal. In a further aspect of the invention, a wafer-contacting pin, such as one of the pins 122,300,400 shown in FIGS. 1, 15, 18, is coated with a relatively hard, electrically conductive film, such as titanium nitride (TiN) or titanium aluminum nitride (TiAlN). The coating provides a relatively hard, abrasion resistant material that enhances the ruggedness of the pin. In the case where the pin is formed from silicon, the TiN coating, for example, is more conductive than the silicon pin such that the likelihood of electrical arcing is reduced in comparison with an uncoated pin. In addition, the coating inhibits so-called wafer bonding in which two silicon surfaces tend to stick together during extreme processing conditions, e.g., relatively high temperatures. It is understood that potentially contaminating particles can be generated when a wafer bond between a wafer and a wafer-contacting pin is broken. The coating can be applied to the pin using a variety of techniques including chemical vapor deposition and reactive sputtering. For chemical vapor deposition to provide a TiN coating, an exemplary precursor gas is titanium chloride. For reactive sputtering a titanium target can be used and nitrogen gas can be added to an argon gas environment. It is understood that the TiN or TiAlN coating can be applied to cover the entire pin, as well as only targeted portions corresponding to the pin/wafer interface. It is further understood that the TiN coating can be applied in discrete portions or as a continuous coating. The thickness of the coating can vary from about 0.1 micrometers to about 10.0 micrometers, and more preferably from about 2 micrometers to about 5 micrometers. A preferred coating thickness is about 5 micrometers. In a further aspect of the invention, the materials for the various components are selected to provide desired features of the assembly, e.g., mechanical durability; electrical conductivity; and minimal particulation. Exemplary materials for the wafer-contacting pin include silicon and graphite. It is understood that silicon is conductive in its intrinsic state at elevated temperatures. Exemplary materials for the main structural members, the retainer members, and the bias member include silicon carbide, graphite and vitreous or vacuum impregnated graphite, which can be coated with titanium carbide. The graphite retainer and bias members can be fabricated from graphite sheets using wire electron discharge machine (xe2x80x9cwire EDMxe2x80x9d), laser machining and conventional cutting techniques. The graphite bias and retaining members maintain a steady, i.e., invariant, spring constant over a wide range of temperatures. This allows the wafer holder assembly to be adjusted at room temperature for operation at temperatures of 600xc2x0 C. and higher, which can occur during the ion implantation process. The graphite components also provide a conductive pathway for grounding the wafer, even where insulative sleeves for the bearing members are used. The wafer holder assembly of the present invention provides a structure that withstands the relatively high temperatures and ion beam energies associated with SIMOX wafer processing. In addition, the likelihood of wafer contamination is reduced since the ion beam strikes only silicon thereby minimizing carbon contamination and particle production. Furthermore, the likelihood of the electrical discharge from the wafer is minimized due to the selection of conductive materials/coatings for the assembly components and/or the geometry of the wafer-contacting pins. One skilled in the art will appreciate further features and advantages of the invention based on the above-described embodiments. Accordingly, the invention is not to be limited by what has been particularly shown and described, except as indicated by the appended claims. All publications and references cited herein are expressly incorporated herein by reference in their entirety.
	abstract	Sensors are configured to repeatedly monitor variables of a physical system during its operation. A novelty detection system is responsive to the sensors and is configured to repeatedly observe into an associative memory, states of associations among the variables that are repeatedly monitored, during a learning phase. The novelty detection system is further configured to thereafter observe at least one state of associations among the variables that are sensed relative to the states of associations that are in the associative memory, to identify a novel state of associations among the variables. The novelty detection system may determine whether the novel state is indicative of normal operation or of a potential abnormal operation. Multiple layers of learning for real-time diagnostics/prognostics also may be provided.
047770139	abstract	A high-temperature gas cooled nuclear reactor system comprises a containment building, a concrete reactor pressure vessel inside the containment building, and a safety relief valve connected to the concrete reactor pressure vessel. The spring of the safety valve consists of a material with a spring constant decreasing as temperature rises. A heat exchanger is provided in close proximity to cool the spring of the safety valve which is subject to the heat of the reactor coolant escaping when the safety valve is open. The heat exchanger of the safety valve is connected to a liner cooling system of the concrete reactor pressure vessel.
	abstract	A personal radiation protection garment that substantially contours to an operator's body is suspended from a suspension means. The garment is operable to protect the operator from radiation. The suspension means is operable to apply constant force. The suspension means allows operator wearing protective radiation garment to move freely in the X, Y, and Z spatial planes simultaneously, such that the protective radiation garment is substantially weightless to operator. A radiation protection face shield and flap can also be suspended from suspension means, such that face shield and flap are substantially weightless to operator. The suspension means can be mounted to a ceiling.
054596750	summary	The present invention is concerned generally with a system and method for reliably monitoring industrial processes having nonwhite noise characteristics. More particularly, the invention is concerned with a system and method for removal of nonwhite noise elements or serially correlated noise, allowing reliable supervision of an industrial process and/or operability of sensors monitoring the process. Conventional parameter-surveillance schemes are sensitive only to gross changes in the mean value of a process, or to large steps or spikes that exceed some threshold limit check. These conventional methods suffer from either large numbers of false alarms (if thresholds are set too close to normal operating levels) or a large number of missed (or delayed) alarms (if the thresholds are set too expansively). Moreover, most conventional methods cannot perceive the onset of a process disturbance or sensor deviation which gives rise to a signal below the threshold level for an alarm condition. In another conventional monitoring method, the Sequential Probability Ratio Test ("SPRT") has found wide application as a signal validation tool in the nuclear reactor industry. Two features of the SPRT technique make it attractive for parameter surveillance and fault detection: (1) early annunciation of the onset of a disturbance in noisy process variables, and (2) the SPRT technique has user-specifiable false-alarm and missed-alarm probabilities. One important drawback of the SPRT technique that has limited its adaptation to a broader range of nuclear applications is the fact that its mathematical formalism is founded upon an assumption that the signals it is monitoring are purely Gaussian, independent (white noise) random variables. It is therefore an object of the invention to provide an improved method and system for continuous evaluation and/or modification of industrial processes and/or sensors monitoring the processes. It is another object of the invention to provide a novel method and system for statistically processing industrial process signals having virtually any form of noise signal. It is a further object of the invention to provide an improved method and system for operating on an industrial process signal to remove unwanted serially correlated noise signals. It is still an additional object of the invention to provide a novel method and system utilizing a pair of signals to generate a difference function to be analyzed for alarm information. It is still a further object of the invention to provide an improved method and system including at least one sensor for providing a real signal characteristic of a process and a predicted sensor signal allowing formation of a difference signal between the predicted and real signal for subsequent analysis free from nonwhite noise contamination. It is also an object of the invention to provide a novel method and system wherein a difference function is foraged from two sensor signals, and/or pairs of signals and nonwhite noise is removed enabling reliable alarm analysis of the sensor signals. It is yet an additional object of the invention to provide an improved method and system utilizing variable pairs of sensors for determining both sensor degradation and industrial process status. Other objects, features and advantages of the present invention will be readily apparent from the following description of the preferred embodiments thereof, taken in conjunction with the accompanying drawings described below.
051200295	abstract	A crucible furnace or molten metal transfer vessel is provided with a composite lining having an outer insulating liner and an inner cast working liner. The insulating liner has a high insulating characteristic and the working liner has a high heat-retention characteristic.
047524332	abstract	A vent system for controlling hydraulically actuated drive means for selectively moving pluralities of rod clusters respectively connected to the plurality of drive means between fully inserted positions within the lower barrel assembly of the vessel in telescoping relationship with fuel rod assemblies contained therein, and a fully withdrawn position. Each drive means responds to the reactor coolant fluid pressure and includes a leakage passage and an outlet channel through which a leakage flow may pass under control of the vent system, the latter comprising a valve arrangement, flow restrictors and a common orifice through which the outlet channels are selectively connected to essentially ambient pressure, for establishing a pressure differential within each drive means producing a net force for moving the drive rods and associated clusters to the withdrawn position at which the drive means mechanically latch the fully withdrawn clusters for mechanically supporting same in the fully withdrawn position. The valve arrangement may be selectively actuated to reestablish the pressure differential within each drive means thereby to raise the associated drive rods and release same from the mechanically latched positions, after which the valve arrangement is selectively operated to establish pressure equilibrium within the drive means to permit the corresponding drive rods and associated control rod clusters to fall by force of gravity to the fully inserted positions thereof. Display indications and control panel operation provide for operator sensing of the rod cluster positions and control of the valve arrangement.
	summary
048667447	claims	1. A scattering beam eliminating device for an X-ray CT apparatus including X-ray irradiation means for irradiating a fan-shaped X-ray beam to a patient, the fan-shaped X-ray beam having an X-ray illumination area defined in a fan-out direction and a slice direction, and an X-ray detector having an X-ray entrance surface and X-ray detection elements arranged in the fan-out direction to detect the X-ray incident thereto through the X-ray entrance surface, the detector detecting the X-ray which has penetrated the patient, the scattering beam eliminating device comprising: (A) a scattering beam eliminating member having an X-ray exit surface positioned on the X-ray entrance surface of the X-ray detector and fixed to the X-ray entrance surface, the scattering beam eliminating member including: (i) an array of substantially parallel plates made of an X-ray absorbing material and spaced apart in the slice direction; and (ii) a plurality of X-ray transmission areas, each located in the space between an adjacent pair of said plates and made of an X-ray transmitting material, and (B) a collimator disposed between said X-ray illumination means and said scattering beam eliminating member for defining an X-ray illumination area in a fan-out direction and a slice direction. X-ray irradiation means for irradiating a fan-shaped X-ray beam to a patient, the fan-shaped X-ray beam having an X-ray illumination area defined in a fan-out direction and a slice direction; an X-ray detector, having an X-ray entrance surface, including substantially parallel plate-like X-ray detection elements for detecting X-ray radiation incident thereon said X-ray detection elements being spaced apart in said fan-out direction; a scattering beam eliminating member having an X-ray exit surface positioned on the X-ray entrance surface of the X-ray detector and fixed to the X-ray entrance surface, the scattering beam eliminating member including an array of substantially parallel, plates made of an X-ray absorbing material, spaced apart in the slice direction, and a plurality of X-ray transmission areas located in the space between an adjacent pair of said plates, said plurality of X-ray transmission areas being made of an X-ray transmitting material. 2. A scattering beam eliminating device according to claim 1, in which the X-ray entrance surface of the X-ray detector and said X-ray exit surface of said scattering beam eliminating member are curved in the fan-out direction and have substantially the same curvature. 3. A scattering beam eliminating device according to claim 1, in which said plates are made of a metal selected from the group consisting of lead, molybdenum and tungsten. 4. A scattering beam eliminating device according to claim 1, in which said X-ray transmission areas are made of aluminum. 5. A scattering beam eliminating device according to claim 1, in which each of said plates is 30 to 80 .mu.m in thickness and said plates are spaced apart with a pitch of 100 to 200 .mu.m. 6. An X-ray CT apparatus comprising:
	claims	1. An imaging apparatus which has a movable element related to imaging and an image sensing element, and has a function of sensing an image of an object with the image sensing element and reading as an image signal a signal generated by the image sensing element, comprising:a control unit arranged to stop movement of the element related to imaging, and, after stopping the movement elapse of predetermined time from stopping movement control for the element, starting to start reading of a signal generated by the image sensing element. 2. The apparatus according to claim 1, wherein the element related to imaging is a grid arranged between the object and the image sensing element. 3. The apparatus according to claim 1, wherein said apparatus further comprises an irradiation detection unit arranged to detect irradiation for the object, and said control unit controls the stopping of movement of the element related to imaging on the basis of a detection result from said irradiation detection unit. 4. The apparatus according to claim 1, wherein after stopping movement of a grid, said control unit starts reading the signal from the image sensing element after an elapse of a predetermined time. 5. The apparatus according to claim 4 1, wherein said control unit determines in advance the predetermined time on the basis of at least one of an irradiation time for the object and a moving speed of the element related to imaging. 6. The apparatus according to claim 1, wherein said apparatus further comprises a vibration detection unit arranged to detect a vibration state of the image sensing element due to movement of the element related to imaging, andsaid control unit controls a start of reading an accumulated signal from the image sensing element on the basis of a detection result from said vibration detection unit. 7. The apparatus according to claim 1, wherein irradiation for the object includes radiation irradiation. 8. An imaging apparatus which has a movable element related to imaging and an image sensing element, and has a function of sensing an image of an object with the image sensing element and reading as an image signal a signal generated by the image sensing element, comprising:drive unit arranged to move the element related to imaging by the image sensing element; andcontrol unit arranged to control said drive unit to operate the element related to imaging at a predetermined speed without any acceleration during an operation period related to reading a signal from the image sensing element. 9. The apparatus according to claim 8, wherein the element related to imaging is a grid inserted between the object and the image sensing element. 10. The apparatus according to claim 8, wherein irradiation for the object includes radiation irradiation. 11. The apparatus according to claim 10, wherein the radiation comprises X-rays. 12. An imaging apparatus which has a movable element related to imaging and an image sensing element, and has a function of sensing an image of an object with the image sensing element and reading as an image signal a signal generated by the image sensing element, comprising:drive unit arranged to move the element related to imaging; andcontrol unit arranged to control said drive unit to operate the element related to imaging at a uniform acceleration during an operation period related to reading a signal from the image sensing element. 13. The apparatus according to claim 12, wherein the element related to imaging is a grid inserted between the object and the image sensing element. 14. The apparatus according to claim 12, wherein irradiation for the object includes radiation irradiation. 15. The apparatus according to claim 14, wherein the radiation comprises X-rays. 16. An imaging apparatus which has a movable element related to imaging and an image sensing element, and has a function of sensing an image of an object with the image sensing element and reading as an image signal a signal generated by the image sensing element, comprising:drive unit arranged to move the element related to imaging; andcontrol unit arranged to control execution of a drive operation related to image acquisition upon determining that a value of a vibration is not more than a predetermined value during an operation period related to an image read from the image sensing element. 17. The apparatus according to claim 16, wherein the element related to imaging is a grid inserted between the object and the image sensing element. 18. The apparatus according to claim 16, wherein irradiation for the object includes radiation irradiation. 19. The apparatus according to claim 18, wherein the radiation comprises X-rays. 20. An imaging apparatus having a function of sensing an image of an object with an image sensing element and reading as an image signal a signal generated by the image sensing element, comprising:a drive unit arranged to move the image sensing element; anda control unit arranged to stop movement of the image sensing element by said drive unit, and, after stopping the movement elapse of a predetermined time from stopping movement control for the element, starting to start reading of an accumulated signal from the image sensing element. 21. The apparatus according to claim 20, wherein after stopping movement of the image sensing element, said control unit starts reading the signal from the image sensing element after an elapse of a predetermined time. 22. The apparatus according to claim 20, wherein said apparatus further comprises vibration detection unit arranged to detect a vibration state of the image sensing element, andsaid control unit controls a start of reading of the signal from the image sensing element on the basis of a detection result from said vibration detection unit. 23. The apparatus according to claim 20, wherein irradiation for the object includes radiation irradiation. 24. An imaging apparatus having a function of sensing an image of an object with an image sensing element and reading as an image signal a signal generated by the image sensing element, comprising:drive unit arranged to move the image sensing element; andcontrol unit arranged to control said drive unit to operate the image sensing element at a predetermined speed without any acceleration during an operation period related to reading a signal from the image sensing element. 25. The apparatus according to claim 24, wherein irradiation for the object includes radiation irradiation. 26. The apparatus according to claim 25, wherein the radiation comprises X-rays. 27. An imaging apparatus having a function of sensing an image of an object with an image sensing element and reading as an image signal a signal generated by the image sensing element, comprising:drive unit arranged to move the image sensing element; andcontrol unit arranged to control said drive unit to operate the image sensing element at a uniform acceleration during an operation period related to reading a signal from the image sensing element. 28. The apparatus according to claim 27, wherein irradiation for the object includes radiation irradiation. 29. The apparatus according to claim 28, wherein the radiation comprises X-rays. 30. An imaging apparatus having a function of sensing an image of an object with an image sensing element and reading as an image signal a signal generated by the image sensing element, comprising:drive unit arranged to move the image sensing element; andcontrol unit arranged to control execution of a drive operation related to image acquisition upon determining that a value of a vibration is not more than a predetermined value during an operation period related to an image read from the image sensing element. 31. The apparatus according to claim 30, wherein irradiation for the object includes radiation irradiation. 32. The apparatus according to claim 31, wherein the radiation comprises X-rays. 33. An imaging method of sensing an image of an object with an image sensing element and reading a signal generated by the image sensing element while moving a movable element related to imaging, comprising:stopping movement of the element related to imaging, and, after stopping the movement elapse of a predetermined time from stopping movement control for the element, starting reading of a signal from the image sensing element. 34. An imaging method of sensing an image of an object with an image sensing element and reading a signal generated by the image sensing element while moving a movable element related to imaging, comprising:in moving the element related to imaging at the time of image sensing by the image sensing element, controlling operation of the element related to imaging at a predetermined speed without any acceleration during an operation period related to reading of a signal from the image sensing element. 35. An imaging method of sensing an image of an object with an image sensing element and reading a signal generated by the image sensing element while moving a movable element related to imaging, comprising:in moving the element related to imaging at the time of image sensing by the image sensing element, controlling operation of the element related to imaging at a uniform acceleration during an operation period related to reading a signal from the image sensing element. 36. An imaging method of sensing an image of an object with an image sensing element and reading a signal generated by the image sensing element while moving a movable element related to imaging, comprising:in moving the element related to imaging at the time of image sensing by the image sensing element, controlling execution of a drive related to image acquisition upon determining that a value of a vibration of the image sensing element is not more than a predetermined value during an operation period related to an image read from the image sensing element. 37. An imaging method of sensing an image of an object with a movable image sensing element and reading a signal generated by the image sensing element, comprising:stopping movement of the image sensing element, and, after stopping the movement elapse of a predetermined time from stopping movement control for the element, starting reading of a signal from the image sensing element. 38. An imaging method of sensing an image of an object with a movable image sensing element and reading a signal generated by the image sensing element, comprising:controlling operation of the image sensing element at a predetermined speed without any acceleration during an operation period related to reading a signal from the image sensing element. 39. An imaging method of sensing an image of an object with a movable image sensing element and reading a signal generated by the image sensing element, comprising:controlling operation of the image sensing element at a uniform acceleration during an operation period related to reading a signal from the image sensing element. 40. An imaging method of sensing an image of an object with a movable image sensing element and reading a signal generated by the image sensing element, comprising:controlling execution of a drive operation related to image acquisition upon determining that a value of a vibration of the image sensing element is not more than a predetermined value during an operation period related to an image read from the image sensing element. 41. A computer-readable storage medium wherein said storage medium stores a processing program for executing said imaging method of claim 33. 42. A computer-readable storage medium wherein said storage medium stores a processing program for executing said imaging method of claim 34. 43. A computer-readable storage medium wherein said storage medium stores a processing program for executing said imaging method of claim 35. 44. A computer-readable storage medium wherein said storage medium stores a processing program for executing said imaging method of claim 36. 45. A computer-readable storage medium wherein said storage medium stores a processing program for executing said imaging method of claim 37. 46. A computer-readable storage medium wherein said storage medium stores a processing program for executing said imaging method of claim 38. 47. A computer-readable storage medium wherein said storage medium stores a processing program for executing said imaging method of claim 39. 48. A computer-readable storage medium wherein said storage medium stores a processing program for executing said imaging method of claim 40.
039490262	summary	This invention relates to a method of fabrication of nuclear reactor fuel elements which are primarily intended for use in high-temperature reactors. In general, the fuels currently employed in reactors of this type are made up of spherical particles of uranium oxide or uranium carbide which are bonded together by a mixture of graphite and thermosetting resin. Fuels of this type are placed within a graphite sheath or jacket. The methods of fabrication of these elements are attended by a number of drawbacks, in particular the need to subject the fuel to high pressures in order to permit polymerization of the resin and shaping of the element. Furthermore, a substantial temperature difference is developed between the jacket and the graphite at the time of operation of said elements. Among the different methods of fabrication of fuel elements which have been known up to the present time, mention can be made of the method described in French Pat. No. 1,588,611 filed on Sept. 23, 1968 by Commissariat a l'Energie Atomique, in which a graphite jacket obtained by extrusion is filled with fuel particles coated with a layer of graphite by spraying a mixture containing the graphite and an organic diluent. The practical application of the method according to the present invention is more straightforward. The high degree of porosity of the raw fuel element is such that the gas can readily penetrate throughout the fuel element. Finally, the method permits the fabrication of fuel elements of substantial length (lengths of 50 to 100 cm can readily be achieved). The method according to the invention meets technical requirements more effectively than those referred-to in the foregoing, particularly insofar as fuel elements having good characteristics can accordingly be obtained by virtue of the initial porosity of the semi-finished products. The method under consideration is characterized in that a jacket is fabricated by moulding in the dry state and without application of pressure a mixture of graphite and powdered resin and said jacket is subjected to a coking treatment; said jacket is then filled with a mixture of fissile particles and of graphite; the complete assembly thus obtained is then closed and treated with a gaseous hydrocarbon. According to one advantageous feature of the invention, the treatment which consists in coking the jacket is followed by a treatment of impregnation with a stream of gaseous hydrocarbon prior to filling of said jacket with the fuel particles. This treatment permits better consolidation of the jacket. According to another feature of the invention, the resin used as a constituent of the mixture which is intended to form the jacket is a phenolic resin. The method according to the invention is carried out in the following manner. A mixture of graphite having a suitable particle size and of powdered resin is placed in an aluminium mould. The mould is heated in an oven to 300.degree.C in order to obtain polymerization of the mixture. The jacket which is thus obtained is then heated to 900.degree.C in a nitrogen atmosphere in order to ensure that coking of the jacket is carried out. This coking treatment can be extended by impregnation of the jacket at the same temperature in a stream of gaseous hydrocarbon such as methane, for example. At this stage, the majority of the pores of the jacket have a diameter within the range of 30 to 50 .mu.. Said jacket is then filled with a selected mixture of fuel particles and of graphite; the particle size distribution of said graphite is comprised between 0 and 125 .mu. and can have different values for the same mixture. This mixture is carried out in the dry state or in wet phase, for example in an alcohol. In order to facilitate the obtainment of a fuel mixture having contiguous particles, a solid or hollow core having the same composition as the jacket can be placed at the center of this latter if so required. In the case of a given ratio of fissile material to graphite, the diameter of the core is chosen so as to have particles which are in close contact with each other. After the fuel mixture has been placed within the jacket, said mixture is subjected to light vibrations without application of pressure. The jacket thus obtained is then closed by a plug formed from a paste which exhibits a low degree of shrinkage at the time of baking. This paste consists of graphite and of an aqueous binder. The fuel element thus constituted is then heated to a temperature within the range of 700.degree.C and 1000.degree.C in a stream of gaseous hydrocarbon (such as methane, propane, propene, for example). The method according to the invention makes it possible to obtain fuel elements of homogeneous density within the range of 1.5 to 2 without any marked discontinuity between the jacket and the fuel. Moreover, the raw fuel elements considered as semi-finished products, that is to say prior to densification, which are obtained by means of the method according to the invention have a uniformly distributed porosity. This permits uniform impregnation of the element with the gaseous hydrocarbons without any filling of pores with excessive deposits of pyrolytic carbon. Moreover, the total residual coke yield remains of low value, namely of the order of 3 to 5 %. Since the jacket is obtained by moulding, and depending on the application which is contemplated, it is an easy matter to fabricate fuel elements having centering fins, external cooling fins, internal heat distribution fins, or either a solid or hollow core.
	summary
044180357	summary	BACKGROUND OF THE INVENTION This invention relates in general to monitoring of the cooling action and condition of coolant in a nuclear power reactor. A required safety measure for operation of nuclear power plants involves detection of coolant level in the reactor fuel core, to provide warnings of dangerous coolant loss. The installation of measuring and monitoring systems for such purposes is regarded as very costly because of equipment costs and modification of reactor design to accommodate such equipment. Further, providing only coolant level information is sometimes insufficient for recognition of inadequate fuel core cooling caused by significant reactor malfunction such as high void fraction-pumped flow and stagnant boil off. On the other hand, non-significant coolant loss detection must be avoided to prevent unnecessary and wasteful power plant shutdowns. It is, therefore, an important object of the present invention to monitor coolant conditions in an economically feasible and meaningful manner. An additional object is to provide advanced warnings of inadequate fuel core cooling by a monitoring system covering a full range of operation. SUMMARY OF THE INVENTION In accordance with one embodiment of the present invention, loss of coolant in the fuel core of a nuclear reactor may be detected through a simple rearrangement of local power rate sensors of the type disclosed in prior application Ser. No. 888,881 filed Mar. 21, 1979, now U.S. Pat. No. 4,298,430 owned in common with the present application by the same assignee. In this type of sensor, the gamma radiation heated sensor body is designed to exhibit a varying temperature distribution because of heat transfer to the coolant within the fuel core. Such sensors include double junction thermocouples at each of a plurality of vertically spaced local measurement zones. The tip portion of the thermocouple at which the faster acting junction is located, is vertically positioned above the other junction so that the faster acting junction normally acts as the cold junction spaced above the hot region of the sensor core as long as reactor coolant is in thermal contact with the sensor throughout. When a drop in coolant level occurs below the faster acting junction, however, the faster acting junction then acts as the hot junction. Thus, depletion of the coolant manifested by a drop in coolant level below the faster acting junction will produce a reversal in polarity of the differential signal voltage across the junctions. Such a reversal in polarity may be used to trigger an alarm or initiate an action sequence in the event of a reactor accident resulting in coolant loss. In accordance with another embodiment of the invention, an electrical heating device is embedded within the sensor body amongst the double junction thermocouples to increase the level of internal heating above that produced by gamma radiation. The level detection operation is thereby enhanced and data obtained by the signal output of the sensors for determining the heat transfer characteristics of the external surface of the sensors at critical levels. An important aspect of the invention resides in the aforementioned modification of gamma sensors so as to render them operative for multiple function monitoring purposes, including the monitoring of coolant level, determination of heat transfer coefficients, and temperature surveillance as well as to monitor local power distribution. The advantages and benefits of the invention are further enlarged by extension of the sensors into the upper dome of a nuclear reactor vessel in order to monitor coolant conditions within the dome. An economical system may thereby be designed for providing safety monitoring facilities based on existing sensors for monitoring local power distribution.
	summary
055047882	abstract	A device for inspecting the support plates and tube sheet of a nuclear steam generator. The device comprises a boom for extending into an access port of a steam generator and into a lane separating two rows of members, the boom being uprightable within the lane; and a video camera attached to the boom for inspecting the tube members and support plates within the lane when the boom is uprighted. In a preferred embodiment the video camera comprises a charge coupled device affixed to a remotely controlled pan and tilt mechanism.
050193252	summary	BACKGROUND OF THE INVENTION This invention generally relates to tooling for removing and installing a control rod drive from the drive housing mounted in the vessel of a boiling water reactor, and is specifically concerned with a compact and lightweight installation and removal assembly capable of expeditiously and remotely performing a control rod drive installation and removal operation without the necessity of providing special support structures in the undervessel cavity. Tooling systems and methods for removing and installing the control rod drives from the drive housings of boiling water reactors are known in the prior art. Some of these tooling systems include a truck or carriage to which a beam is pivotally mounted. When the beam is swung into a horizontal position, the combination of the carriage and beam can be rolled along the service rails normally present in the undervessel cavity located beneath the reactor vessel. The beam is provided with a bucket for capturing an end of a control rod drive assembly, as well as a lifting and lowering mechanism for moving this bucket up or down when the beam is in a vertical position beneath a control drive housing. Examples of such tooling systems are disclosed in U.S. Pat. Nos. 4,288,290, 4,292,133, and Japanese patent 29,596. While pivoting-beam type tooling systems have met with some success in installing and removing the control drive rods of boiling water reactors, the applicants have observed that each of these prior art systems has a number of operational shortcomings. However, before these shortcomings can be appreciated, some background as to the environment where these tooling systems are used is necessary. Boiling water reactors generally include a cylindrically shaped reactor vessel which is supported over a cylindrically-shaped concrete room called the undervessel cavity in the art. Extending down from the bottom of the reactor vessel is an array of tube-like housings for housing the control rod drives that slide control rods up and down within the fuel assemblies disposed within the reactor vessel in order to control the fission reaction which occurs therein. Over a period of time (which is typically approximately four years) the bushings and seals of the control rod drives begin to wear out, thereby necessitating their replacement. The principle purpose of the undervessel cavity disposed beneath the reactor vessel is to provide access to the control rod drives and other reactor components extending downwardly from the bottom of the reactor vessel so that they may be serviced. Such undervessel cavities are typically provided with a pair of service rails which allow maintenance equipment to be easily shuttled across the diameter of the cylindrically-shaped undervessel cavity. To allow such maintenance equipment to be positioned at any given point under the reactor vessel, the ends of these service rails include wheels which engage a circular track that circumscribes the inner wall of the undervessel cavity. Hence a maintenance device may be moved in a polar-coordinate fashion under the reactor vessel by traversing the device to a selected point along the service rails and by rotating these service rails from zero to 360 degrees until the device is disposed under the housing of a selected control rod drive or other component. Unfortunately, the undervessel cavity provides very little clearance for the entrance and operation of control rod drive installation and removal systems. While the bottom ends of the housings for the control rod drives are almost seven feet from the top of the service rails, the actual usable clearance is often only about four feet above the service rails due to the large number of delicate instrument tubes which extend from the bottom of the reactor vessel, and further due to the "forest" of electrical cables used to power the control monitors and control rod drives which drape down from the bottom of the vessel. The applicants have noted that the tooling systems developed thus far for the removal and installation of such control rod drives suffer from a number of deficiencies which could bear improvement. These systems must be manually wheeled out onto the service rails, thus exposing workers to the "shine" of radiation emitted by the reactor vessel. Some of these systems use chain and sprocket drive mechanisms for elevating the control rod drives into position which can damage or completely cut through any of the maze of instrumentation tubes and electrical cables which hang down from the bottom of the reactor vessel. The operation of such tooling systems must be very carefully monitored by maintenance personnel standing in the immediate proximity to insure that none of the moving chains and sprockets damages any of the reactor components. The long chains such systems are further prone to stretching, which makes the automatic operation of these machines difficult as the number of sprocket turns necessary to elevate a particular control rod drive can vary. Others of these tooling systems are multi-component systems which include separate control rod elevating mechanisms or bolt removal assemblies that necessitate the installation of special tracks within the undervessel cavity. Some of these systems are considerably heavier than the existing service rails can carry, thereby necessitating replacing these rails. The installation of additional tracks and the replacement of the existing service rails again adds substantiallY to the time required to remove and replace worn control drive housings. Further, the pivotal stroke of the beams of these systems is very often larger than the clearance afforded within the undervessel cavity at a given drive housing, which necessitates manually moving the carriage of the device as the carrying beam is pivoted to avoid mechanical interference between the ends of the pivoting beam and one or more of the instrument tubes, electrical cables and other reactor components. Such manual positioning and repositioning of the carriage on the service rails greatly protracts the operational time required to either remove or install a control rod drive, which has the effect of requiring the maintenance personnel operating the system to spend substantial amounts of time in the radioactive undervessel cavity. Finally, the control rod drive lifting mechanisms associated with the pivoting beams provide no means for facilitating a rapid alignment between a control rod drive and a particular housing, and additionally are not completely reliable in operation. All of these are significant drawbacks that necessitate a great deal of manual labor in a highly radioactive environment. Clearly, there is a need for a control rod drive installation and removal system that is sufficiently lightweight and compact in structure so that is may be used solely in conjunction with the service rails already provided in the undervessel cavity, and whose operational movements are short and directed either within or under the carriage of the system so as to avoid mechanical interference with the reactor components. Moreover, the system should be automatically and remotely operable, and the pivoting stroke of the beam of the tool should be short enough to eliminate or at least minimize the necessity for multiple movements of the carriage along the service rails whenever the support beam is pivoted. Ideally. such a system should further provide a self-contained lifting and lowering mechanism which is capable of moving a control rod drive from a position at the bottom end of the carriage to a position completely installed within a drive housing without the necessity of adding additional elevating mechanisms to the cradle. Finally, such a system should have a means for facilitating the rapid alignment of the end of a control rod drive with the open end of a drive housing so as to expedite the operation of the system and to minimize the exposure of the system operators to potentially harmful radiation. SUMMARY OF THE INVENTION The invention is both a system and method for removing and installing a control rod drive which overcomes the shortcomings associated with the prior art and which may be substantially remotely and automatically operated. The system generally comprises a control rod drive installation and removal assembly including a carriage that is remotely and precisely movable to a specific location along the service rails disposed in the undervessel cavity of the reactor, and a cradle pivotally connected to the carriage and having a length less than the length of a control rod drive in order to minimize the possibility of mechanical interference between the installation and removal assembly and the drive housing, instrument tubes, and electrical cables hanging down from the bottom of the vessel. Moreover, the cradle includes a lifting and lowering mechanism which advantageously has a working stroke long enough to move the control rod drive from between a position where its bottom end is coterminous with the bottom end of the cradle, and a position where it is installed within a drive housing. The ability of the lifting and lowering mechanism to support and extend the control drive well beyond the upper end of the cradle greatly expedites the operation of the assembly by obviating the need for positioning separate elevating tools under the control rod drive. The carriage includes a plurality of pairs of wheels that rollingly engage the service rails, and a drive train including a precisely controllable and lightweight hydraulic motor coupled to a simple transmission formed from a driven sprocket, and two drive sprockets connected to two of the cradle wheels. In the preferred embodiment, the two, forward most carriage wheels that engage the an angle iron projecting from the track of the service rails are driven by the hydraulic motor to assure a positive traction at all times between the carriage and the service rails. The lifting and lowering mechanism of the cradle includes a leadscrew drive assembly for moving a control rod drive from a position within the cradle to a position coterminous with the top end of the cradle, and a hydraulic cylinder for moving the control rod drive beyond the top end of the cradle to an installed position within a drive housing. To prevent the leadscrew drive assembly and the hydraulic cylinder from mechanically interfering with one another, the hydraulic cylinder is slidably connected to the cradle so that it is movable between a recess within the cradle when not in use, to a position directly beneath and in tandem with the control rod drive when in use. In the preferred embodiment, the slidable connection is effected by means of a toggle linkage actuated by means of a kick-out cylinder. The cradle further includes a securing mechanism for detachably securing a control rod drive to its underside. In the preferred embodiment, the securing assembly is formed from a retaining bracket capable of capturing one end of a control rod drive, and an extendable and retractable pair of jaws for capturing the control rod drive at another point. The jaws include rollers so that the control rod drive may freely slide through the Jaws when the control rod drive is moved vertically by either the leadscrew drive assembly or the hydraulic cylinder of the lifting and lowering mechanism. Moreover, both the retaining bracket and the extendable and retractable jaws compliantly secure the control rod drive in order to assist the system operator in aligning an end of the control rod drive into a tubular drive housing. Finally, the retaining bracket includes an opening to allow the piston rod of the hydraulic cylinder of the lifting and lowering mechanism to extend through the bracket and to lift the end of the control rod drive secured by the retaining bracket beyond the bracket and into an installed position within a drive housing. The installation and removable assembly further comprises a pivot drive formed by a single hydraulic cylinder mounted between one end of the carriage and one end of the cradle for pivoting the cradle from a vertical to a horizontal position within the undervessel. In the preferred embodiment, the pivot joint between the carriage and the cradle is selected so that the end of the cradle which extends upwardly when the cradle is pivoted is well below the drive housings, instrument tubes and hydraulic power lines so as to avoid mechanical interference therewith. Moreover, the hydraulic cylinder of the pivot drive is connected between one end of the carriage and the relatively short end of the cradle that extends upwardly when the cradle is pivoted vertically. Such an arrangement advantageously allows the hydraulic cylinder to pivot the cradle by means of a relatively short horizontal stroke which is adjacent to the service rails in the undervessel in a position extremely unlikely to mechanically interfere with any of the components of the nuclear vessel. To further minimize any chance of such mechanical interference, the length of the carriage is rendered shorter than the length of the cradle, and the pivot point between the carriage and the cradle is selected so that the length of the carriage is entirely subsumed within the length of the cradle when the cradle is pivoted into a horizontal position. Further adding to the compact dimensions and interference free operation of the assembly is the use of a transfer cart on the lower pair of tracks provided by the service rails in the undervessel cavity. The transfer cart used in the system of the invention includes a raising and lowering mechanism for raising and lowering a control rod drive onto the securing mechanism located on the underside of the cradle. The use of such a transfer cart on these lower tracks obviates the need for the construction and installation of additional rail structures in the undervessel. In short, the system of the invention is capable of both removing and installing a control rod drive from a drive housing mounted in the vessel of a boiling water reactor by means of a single, rapidly-operated tool whose dimensions and operational movements minimize the possibility of mechanical interference with the surrounding reactor components. Moreover, these compact tool dimensions and operational movement eliminate or at least minimize the amount of positioning and repositioning the carriage of the tool must make on the service rails when installing or removing selected control rod drives, which further expedites the overall operation of the tool. These features, in combination with the compliant manner in which the securing means secures the control rod drive as it is lifted by the lifting and lowering mechanism, help to render the system largely remotely controllable, which in turn minimizes the amount of time the maintenance personnel must spend in the radioactive undervessel cavity.
	description	This invention was made with government support under Contract No. DE-NE0008222 awarded by the Department of Energy. The U.S. Government has certain rights in this invention. The invention relates to fuels for nuclear reactors, and more particularly to methods of improving corrosion resistance of nuclear fuels. Enhancing the safety and performance of light water reactors is an ongoing subject of research. Uranium Nitride (UN) and Uranium Silicide (U3Si2) fuels have been selected as the leading candidates for advanced light water reactor fuel due to their high thermal conductivity and density. One of the major weaknesses for UN fuel, however, is its interaction with water and steam at normal operating conditions and high temperatures. The reaction of U3Si2 with water and steam is less severe than UN but any improvement is beneficial. It has been reported in the literature that the grain boundaries of UN and U3Si2 are preferentially attacked when exposed to water or steam and appear to be the major cause for rapid reaction and disintegration of these fissile materials. UN and U3Si2 are fuels with much improved thermal conductivity and density compared to most fuel types. If the water and steam corrosion resistance problem can be solved for UN and U3Si2, it will become a much more attractive accident tolerant fuel pellet. A method for adding additives or dopants to Uranium Nitride (UN) and Uranium Silicide (U3Si2) pellets to improve their water corrosion resistance in nuclear reactor coolant during operation and in high temperature steam in loss of coolant accidents conditions is described. It is found that UN pellets have minimal oxidation resistance in water and steam even at 200° C. U3Si2 has better oxidation resistance than UN but still reacts with air, water, or steam at temperatures higher than 360° C. See E. Sooby Wood, et. al, “Oxidation behavior of U—Si compounds in air from 25 to 1000° C.”, Journal of Nuclear Materials, 484 (2016) pp. 245-257. A method for improving corrosion resistance of nuclear fuels is described herein which includes mixing a powdered fissile material selected from the group consisting of UN and U3Si2 with an additive selected from oxidation resistant materials wherein the powdered fissile material comprises grains having grain boundaries, pressing the mixed fissile and additive materials into a pellet, and sintering the pellet to a temperature greater than the melting point of the additive, sufficient for melting the additive for coating the grain boundaries of the fissile material and densifying the pellet. The additive selected may also have a median particle size significantly lower than the median particle size of the UN or U3Si2 while having a melting point greater than the UN or. In various aspects, the oxidation resistant additive may have a melting point lower than the sintering temperature of the fissile material. In various aspects, when the melting point of the oxidation resistant additive is greater than the sintering temperature of UN or U3Si2, the oxidation resistant particles can have a median particle size less than 10% that of the UN or U3Si2. By the method described herein, small amounts of oxidation resistant compound(s) (less than 20 wt %) are incorporated into the fissile material, (i.e., UN and U3Si2) at the grain boundaries of the material and thus achieve improved oxidation resistance. The additives may be in powder form and may be added or mixed with U3Si2 or UN powders before pressing into pellets and sintering. The additives may be coated to the U3Si2 or UN powders to form protective layers before pressing into pellets and sintering. The oxidation resistant particles may also be applied through vapor deposition (such as physical vapor deposition, chemical vapor deposition, and atomic layer deposition) to green (unsintered) pellets of UN or U3Si2 to coat the outside of the pellet and penetrate into the green pellet as the green pellet has a lot of open pores/channel through the pellet. Upon sintering, the oxidation resistant material will be incorporated into the outside grain structure (grain boundary) of the UN or U3Si2. In certain aspects, the additives include one or a mixture of molybdenum, titanium, aluminum, chromium, thorium, copper, nickel, manganese, tungsten, niobium, zirconium, yttrium, cerium, magnesium, alloys containing at least 50 atomic % of at least one of molybdenum, titanium, aluminum, chromium, thorium, copper, nickel, manganese, tungsten, niobium, zirconium, yttrium, cerium, and magnesium, magnesium nitride, ZrSi2, ZrSiO4, CrSi2, BeO, and UO2 and glassy materials, such as a borosilicate glass. Either the additive or mixture of additives to be mixed with the fissile material has a lower melting point than UN or U3Si2 or, the additive or mixture of additives and the nuclear fuel form low melting point eutectics. In various aspects, densification is achieved via liquid phase sintering or co-sintering. Pellets can be sintered, for example, by using sintering methods selected from the group consisting of pressureless sintering, hot pressing, hot isostatic pressing, spark plasma sintering, field assisted sintering, or flash sintering. Those skilled in the art will recognize that any suitable known sintering method may be used. Also described herein is a nuclear fuel comprising a pellet comprised of compressed and densified grains of a fissile material selected from the group consisting of UN and U3Si2, and an oxidation resistant additive, preferably present in amounts less than about 20% by weight of the fissile material, that coats at least a portion of the grain boundaries of the fissile material. In certain aspects, the additives include one or a mixture of molybdenum, titanium, aluminum, chromium, thorium, copper, nickel, manganese, tungsten, niobium, zirconium, yttrium, cerium, magnesium, alloys containing at least 50 atomic % of at least one of molybdenum, titanium, aluminum, chromium, thorium, copper, nickel, manganese, tungsten, niobium, zirconium, yttrium, cerium, and magnesium, magnesium nitride, ZrSi2, ZrSiO4, CrSi2, BeO, and UO2 and glassy materials, such as a borosilicate glass. As used herein, the singular form of “a”, “an”, and “the” include the plural references unless the context clearly dictates otherwise. Thus, the articles “a” and “an” are used herein to refer to one or to more than one (i.e., to at least one) of the grammatical object of the article. By way of example, “an element” means one element or more than one element. In the present application, including the claims, other than where otherwise indicated, all numbers expressing quantities, values or characteristics are to be understood as being modified in all instances by the term “about.” Thus, numbers may be read as if preceded by the word “about” even though the term “about” may not expressly appear with the number. Accordingly, unless indicated to the contrary, any numerical parameters set forth in the following description may vary depending on the desired properties one seeks to obtain in the compositions and methods according to the present disclosure. At the very least, and not as an attempt to limit the application of the doctrine of equivalents to the scope of the claims, each numerical parameter described in the present description should at least be construed in light of the number of reported significant digits and by applying ordinary rounding techniques. Further, any numerical range recited herein is intended to include all sub-ranges subsumed therein. For example, a range of “1 to 10” is intended to include any and all sub-ranges between (and including) the recited minimum value of 1 and the recited maximum value of 10, that is, having a minimum value equal to or greater than 1 and a maximum value of equal to or less than 10. Methods are described herein for increasing the oxidation resistance of the grain boundaries of fissile materials, such as UN and U3Si2 nuclear fuels, so that interactions with water or steam can be suppressed and the washout of fuel pellets can be minimized should a leak in the fuel rod occur. Improving the oxidation resistance of the grain boundaries will improve the resistance to the oxidation reaction and the resulting fragmentation as the less dense UO2 is formed, which is the key degradation mechanism when U3Si2 and UN are exposed to water or steam at higher temperatures. The grain boundary modification could be the most effective method to improve the corrosion and oxidation resistance of high density fuels like U3Si2 and UN. A method for improving the oxidation resistance of the grain boundaries and improving the corrosion resistance of nuclear fuels is described herein. The method includes mixing a powdered fissile material selected from the group consisting of UN and U3Si2 with an additive selected from oxidation resistant materials. In various aspects, the additives may have a melting point lower than the sintering temperature of the fissile material. The mixture is pressed into a pellet, then sintered to a temperature greater than the melting point of the additive. As the additive melts, it flows around the still solid grains of the fissile material, coating the grain boundaries of the fissile material and densifying the pellet. In various aspects, when the melting point of the oxidation resistant particles is greater than the sintering temperature of UN or U3Si2, the oxidation resistant particles can have a median particle size less than 10% that of the UN or U3Si2. In certain aspects, the additives may be coated to the U3Si2 or UN powders to form protective layers before pressing into pellets and sintering. The oxidation resistant particles may also be applied through vapor deposition (such as physical vapor deposition, chemical vapor deposition, and atomic layer deposition) to green (unsintered) pellets of UN or U3Si2 to coat the outside of the pellet and penetrate into the green pellet as the green pellet has a lot of open pores/channels through the pellet. Upon sintering, the oxidation resistant material will be incorporated into the outside grain structure (grain boundary) of the UN or U3Si2 pellets. The additives distributed along grain boundaries may stop fission gas releases from the U3Si2 grains such as Xe and Kr, as well as volatile fission products like Iodine. This will result in lower rod internal pressure which improves operating margins and the dry storage. In addition to oxidation resistance, the corrosion resistance phase will also prevent U3Si2 from interacting with cladding, plenum spring, and other rod internal components such as spacers which separate U3Si2 from end plug and plenum springs. The additives may be in powder form and may be added or mixed with U3Si2 or UN powders before pressing into pellets and sintering. The additives may be coated to the U3Si2 or UN powders to form protective layers before pressing into pellets and sintering. The desired characteristics of the additive are that it is an oxidation resistant material and that it has a melting point lower than the melting point of the fissile material, either UN or U3Si2, with which it is mixed; and in various aspects, at least 200° C., and in certain aspects, from 200 to 300° C. lower than the sintering temperature of the fissile material. Alternatively, if the melting point of the oxidation resistant particles is greater than the sintering point of UN or U3Si2, then the oxidation resistant particles can have a median particle size less than 10% that of the UN or U3Si2, and in certain aspects, a median particle size less than 1% that of the UN or U3Si2. In certain aspects, exemplary additives include one or a mixture of molybdenum, titanium, aluminum, chromium, thorium, copper, nickel, manganese, tungsten, niobium, zirconium, yttrium, cerium, magnesium, alloys containing at least 50 atomic % of at least one of molybdenum, titanium, aluminum, chromium, thorium, copper, nickel, manganese, tungsten, niobium, zirconium, yttrium, cerium, and magnesium, magnesium nitride, ZrSi2, ZrSiO4, CrSi2, BeO, and UO2 and glassy materials, such as a borosilicate glass. Either (1) the additive or the mixture of additives have lower melting points than the fissile material (UN or U3Si2) with which it is mixed or (2) the additive or the mixture of additives and the nuclear fuel form low melting point eutectics. For example, the fissile material may be UN and may be mixed with BeO as the additive. BeO has a melting point lower than the sintering temperature of UN. Those skilled in the art will be able to determine the melting points or sintering temperatures of the fissile material and the melting points of the oxidative resistant additives, or determine the melting point eutectics of the selected fissile material and additive, and select, according to the method described herein, the appropriate additive or mixture of additives for mixing with either UN or U3Si2. The fuel and additive mixture may be formed into pellets by compressing the mixture of particles in suitable commercially available mechanical or hydraulic presses to achieve the desired “green” density and strength. A basic press may incorporate a die platen with single action capability while the most complex styles have multiple moving platens to form “multi-level” parts. Presses are available in a wide range of tonnage capability. The tonnage required to press powder into the desired compact pellet shape is determined by multiplying the projected surface area of the part by a load factor determined by the compressibility characteristics of the powder. To begin the process, the mixture of particles is filled into a die. The rate of die filling is based largely on the flowability of the particles. Once the die is filled, a punch moves towards the particles. The punch applies pressure to the particles, compacting them to the geometry of the die. In certain pelleting processes, the particles may be fed into a die and pressed biaxially into cylindrical pellets using a load of several hundred MPa. Following compression, the oxidation resistant particles may also be applied through vapor deposition (such as physical vapor deposition, chemical vapor deposition, and atomic layer deposition) to the green (unsintered) pellets of UN or U3Si2 to coat the outside of the pellet and penetrate into the green pellet as the green pellet has a lot of open pores/channel through the pellet. The pellets are sintered by heating in a furnace at temperatures varying with the material being sintered under a controlled atmosphere, usually comprised of argon. Sintering is a thermal process that consolidates the green pellets by converting the mechanical bonds of the particles formed during compression into stronger bonds and greatly strengthened pellets. Upon sintering, the oxidation resistant material will be incorporated into the outside grain structure (grain boundary) of the UN or U3Si2 pellets. The compressed and sintered pellets are then cooled and machined to the desired dimensions. Exemplary pellets may be about one centimeter, or slightly less, in diameter, and one centimeter, or slightly more, in length. Referring to the FIGURE, the hexagons in Part A represent grains of UN or U3Si2 fuel 12 with grains of an additive 14 mixed with the UN or U3Si2 fuel grains 12. As sintering proceeds and the temperature reaches and, in various aspects, surpasses the melting or softening point of the additive, the additive 14 melts or softens, and as shown in Part B of the FIGURE, flows about the mixture, coating all or at least a portion of the UN or U3Si2 fuel grains 12 on the grain boundaries. If the additive has a higher melting point than the sintering temperature of U3Si2 or UN, the fine particles are sintered into the grain boundaries of the larger U3Si2 or UN grains. The fuel 12 grain boundary coverage by the additive 14 in the FIGURE is an ideal case, and the actual coverage may be lower. It is preferred that the additive phase is interconnected. In various aspects, densification is achieved via liquid phase sintering or co-sintering. Pellets can be sintered, for example, by using sintering methods selected from the group consisting of liquid phase sintering, pressureless sintering, hot pressing, hot isostatic pressing, spark plasma sintering, sometimes referred to as field assisted sintering or pulsed electric current sintering. Those skilled in the art will recognize that any suitable known sintering method may be used. In a typical sintering process for producing nuclear fuel pellets, the pressed powder pellets are heated so that adjacent grains fuse together, producing a solid fuel pellet with improved mechanical strength compared to the pressed powder pellet. This “fusing” of grains results in an increase in the density of the pellet. Therefore, the process is sometimes called densification. In hot isostatic pressing, the compaction and sintering processes are combined into a single step. In various aspects, the sintering may be done by a liquid phase sintering and co-sintering processes, both advanced processing technologies which have not heretofore been used in nuclear fuel manufacturing. In liquid phase sintering, the solid grains are insoluble in the liquid so the liquid phase can wet on the solid phase. This insolubility causes the liquid phase to wet the solid, providing a capillary force that pulls the grains together. At the same time, the high temperature softens the solid, further assisting densification. During heating, the particles sinter. The solid grains rearrange when a melt forms and spreads. Subsequent densification is accompanied by coarsening. The liquid wets and penetrates between the solid grains. See German, R. M., Sun, P. & Park, S. J., J Mater Sci (2009) 44: 1. https://doi.org/10.1007/s10853-008-3008-0. In various aspects, the sintering process may be done using spark plasma sintering, wherein external pressure and an electric field are applied simultaneously to enhance the densification of the pressed powder pellets. A pulsed DC current directly passes through the die, as well as the powder compact. The electric field driven densification supplements sintering with a form of hot pressing, to enable lower temperatures and shorter amount of time than typical sintering. The heat generation is internal, in contrast to the conventional hot pressing, where the heat is provided by external heating elements. Pressureless sintering is a well-known sintering method involving the sintering of a powder compact (sometimes at very high temperatures, depending on the powder) without applied pressure. This avoids density variations in the final pellet, which occurs with more traditional hot pressing methods. In another aspect, the sintering may be done by hot isostatic pressing. In this techniques, powders are usually encapsulated in a metallic or glass container. The container is evacuated, the powder out-gassed to avoid contamination of the materials by any residual gas during the consolidation stage and sealed-off. It is then heated and subjected to isostatic pressure sufficient to plastically deform both the container and the powder. The rate of densification of the powder depends upon the yield strength of the powder at the temperatures and pressures chosen. At moderate temperature the yield strength of the powder can still be high and require high pressure to produce densification in an economic time. The method produces a nuclear fuel comprising a pellet comprised of compressed and densified grains of a fissile material selected from the group consisting of UN and U3Si2, and an oxidation resistant additive, preferably present in amounts less than about 20% by weight of the fissile material, that coats at least a portion of the grain boundaries of the fissile material. In certain aspects, the additives include one or a mixture of molybdenum, titanium, aluminum, chromium, thorium, copper, nickel, manganese, tungsten, niobium, zirconium, yttrium, cerium, magnesium, alloys containing at least 50 atomic % of at least one of molybdenum, titanium, aluminum, chromium, thorium, copper, nickel, manganese, tungsten, niobium, zirconium, yttrium, cerium, and magnesium, magnesium nitride, ZrSi2, ZrSiO4, CrSi2, BeO, and UO2 and glassy materials, such as a borosilicate glass. The present invention has been described in accordance with several examples, which are intended to be illustrative in all aspects rather than restrictive. Thus, the present invention is capable of many variations in detailed implementation, which may be derived from the description contained herein by a person of ordinary skill in the art. All patents, patent applications, publications, or other disclosure material mentioned herein, are hereby incorporated by reference in their entirety as if each individual reference was expressly incorporated by reference respectively. All references, and any material, or portion thereof, that are said to be incorporated by reference herein are incorporated herein only to the extent that the incorporated material does not conflict with existing definitions, statements, or other disclosure material set forth in this disclosure. As such, and to the extent necessary, the disclosure as set forth herein supersedes any conflicting material incorporated herein by reference and the disclosure expressly set forth in the present application controls. The present invention has been described with reference to various exemplary and illustrative embodiments. The embodiments described herein are understood as providing illustrative features of varying detail of various embodiments of the disclosed invention; and therefore, unless otherwise specified, it is to be understood that, to the extent possible, one or more features, elements, components, constituents, ingredients, structures, modules, and/or aspects of the disclosed embodiments may be combined, separated, interchanged, and/or rearranged with or relative to one or more other features, elements, components, constituents, ingredients, structures, modules, and/or aspects of the disclosed embodiments without departing from the scope of the disclosed invention. Accordingly, it will be recognized by persons having ordinary skill in the art that various substitutions, modifications or combinations of any of the exemplary embodiments may be made without departing from the scope of the invention. In addition, persons skilled in the art will recognize, or be able to ascertain using no more than routine experimentation, many equivalents to the various embodiments of the invention described herein upon review of this specification. Thus, the invention is not limited by the description of the various embodiments, but rather by the claims.
060552883	description	DESCRIPTION OF THE PREFERRED EMBODIMENT Referring now to the drawings in detail and in particular to FIG. 1 there is shown a conventional nuclear reactor vessel 10 which may be employed in commercial pressurized water reactors for commercial electric power generation. The reactor vessel 10 generally has a vessel shell 12 with a removable head 14 fastened thereto by a plurality of nuts 16 on studs 18. The vessel shell 12 has at least one reactor inlet nozzle 20 and at least one reactor outlet nozzle 22 in fluid flow communication with a nearby steam generator (not shown). In commercial electric power generating plants, a reactor vessel 10 will normally be coupled with two, three or four steam generators and receive primary coolant from reactor coolant pumps associated with the steam generators. Thus, there will normally be one inlet nozzle 20 and one outlet nozzle 22 for each steam generator. A core barrel 24 is suspended from an internal ledge (not shown) and is radially supported at a lower support plate 26 by radial supports 28 attached to the reactor vessel shell 12. The core barrel 24 supports a plurality of fuel assemblies (illustrated by fuel assembly 36) in a core region 30 which is generally defined by a lower core plate 32 and an upper core plate 34. The core region 30 may support up to 150 or more fuel assemblies surrounded and supported by a baffle assembly 38 including a substantially vertical baffle plate 40 (which may be comprised of a plurality of smaller plates which are welded or bolted together) fastened to generally horizontal former plates 42 by baffle plate/former plate bolts 50. Bolts 50 are shown in FIG. 1 as protruding into the core region 30 for purposes of illustration but Generally do not protrude as shown in commercial reactor vessels 10. The former plates 42 may be welded or bolted to the core barrel 24. The primary coolant fluid flows into the reactor vessel 10 from a reactor coolant pump (not shown) through the reactor vessel inlet 20, downwardly through an annulus know as a "downcomer" between the reactor vessel shell 12 and the core barrel 24, upwardly through holes 46 in the lower plate 26, through the core region 30 into an upper plenum 48, and out through reactor vessel outlet 22 to a steam generator (not shown) and then back to the reactor coolant pump. The principal functions of the recirculating coolant fluid are to absorb heat generated by the fuel assemblies 36 and to cool the reactor vessel 10 and its internal structure. FIGS. 2 and 3 illustrate a baffle plate 40/former plate 42 joint design having a modified baffle/former bolt 50 which permits the primary coolant to flow past a modified bolt head portion 52 to cool the bolt 50 and to wash the undersurface 54 of the bolt head portion 52. The undersurface 54 may be the underside of a flanged section of the head portion 52 (as is shown and is preferable) or may be the undersurface of an unflanged head (as shown in U.S. Pat. No. 4,069,102, although not shown herein) or the undersurface of a separate washer (not shown). The baffle plate 40 has a countersunk hole 56 having a diameter extending to a smaller diameter bolt hole 58, and the former plate 42 has a bolt hole 60 aligned with the baffle plate hole 58. The bolt 50 has a shank 62 extending from the undersurface 54 of the bolt head portion 52 into the aligned bolt holes 58,60 to a threaded distal end 64 which threadedly engages the former plate 42. When the baffle assembly 38 is fastened together, the underside 54 of the bolt head portion 52 seats against countersunk surface 64 of the baffle plate 40 and forms a crevice therewith. The bolt head portion 52 has one or more spaced apart peripheral slots 68 which, together with the baffle plate countersunk surface 64, defines fluid flow passageways such as orifices 70 between the countersunk hole 56 and the baffle plate bolt hole 58. The orifices 70 may be sized to permit the primary coolant to flow into and out of the bolt hole 58 and past the bolt head portion underside 54 so that any steam which may form will be vented, any solids which may tend to deposit near the crevice under the bolt head portions 52 will be washed away and the undersurface 54 of the bolt head portion 52 will be cooled. Alternatively and/or additionally, the fluid flow passageways may be formed by one or more holes through the head portion 52 externally of the shank 62 as illustrated by holes 72. FIGS. 4-9 illustrate other embodiments of the present invention where one or more fluid flow passageways may be machined into the baffle plate 40 and the former plate 42 while the baffle assembly 38 is assembled in the reactor vessel. FIGS. 4-7 illustrate embodiments of the present invention wherein two slots 80,82 and 84,86 are machined into the baffle plate 40 by any suitable means such as electro-machining. The slots 80,82 have the same cross-sectional areas whereas slots 84,86 have different cross-sectional areas for providing different flow rates. The slots 80,82 and slots 84,86 extend to slots 90, 92 and 94,96 (having far sides 91, 93 and 95, 97, respectively) in the former plate 42, which permits the primary coolant to flow from one side of the baffle plate to the other side. These slots are sized, positioned and designed to prevent excessively high hydraulic forces of the flowing primary coolant from impinging upon the adjacent fuel assemblies 36. Advantageously, the slots 90,92 and 94,96 may be cut very narrowly and substantially as long as the shank 62 for cooling the bolts 50. FIGS. 8-9 illustrate another embodiment of the present invention wherein two spaced apart slots 100-102 are machined into the baffle plate 40 for controlling the flow of primary coolant from one side of the baffle plate 40 to the other. Slot 100 extends to a slot 104 (having a far side 105) machined into the former plate 42 whereas slot 102 is only partially machined through the baffle plate 40 (extending to an interior far side 103). FIGS. 8-9 also illustrate a design modification where, in addition to one or more fluid flow passageways illustrated by FIGS. 2-9, a fluid flow passageway extending internally of the head portion 52 and the shank 62, as shown by phantom hole 110, may be formed through the bolt 52 for communicating between the countersunk hole 56 and the bolt hole 60 in the former plate 42. Advantageously, an internal hole 110 may extend toward the distal end 64 of the bolt for washing the threads. However, and as has been noted above, bolts having internal passageways in their shanks are difficult to ultrasonically inspect and the internal passageways in the shank may affect the strength of the bolt. FIGS. 8-9 also illustrate an embodiment wherein elongated (above and below the former plates 42) fluid flow passageways, such as phantom slots 106,108, are machined in the baffle plate 40. Advantageously, such a design permits cooling fluid flow from one side of the baffle plate 40 to the other without having to machine or drill holes through the relatively heavier former plates 42. For example, the baffle plates 40 may be on the order of two inches thick whereas the former plates 42 may be on the order of four inches thick. Also, the cross-sectional areas of the slots 106,108 may differ. FIGS. 10 and 11 illustrates a locking cup 110 which may be employed to engage is the slots 99, 100, 101 or 102 in the baffle plate 40 shown in FIG. 8 in order to prevent loosening of the baffle/former bolts 50. As is shown, the locking cup 110 may be a peripheral bowl-shaped extension 114 integral with or welded to the flange on the bolt head portion 52, which may be wedged or otherwise deformed as shown at 116 and 118 to tightly engage fluid flow slots 100 and 102, respectively. In other embodiments, the locking cup 110 may be separate from the bolt head portion 52 in accordance with previously employed locking cup designs. In addition, the locking cup 110 may be crimped into an irregularly shaped feature of the baffle plate. See, e.g., the locking design of U.S. Pat. No. 4,683,108, which is incorporated by this reference. To backfit a reactor vessel 10, the reactor is taken out of service, the vessel is submerged in a pool, e.g., of refueling water, the top 14 removed, the fuel assemblies 36 removed, and the existing bolts removed. The existing bolts may then be replaced by modified bolts 50 illustrated in FIGS. 2-4 and, if preferred, locked on with a locking cup. In some cases it may be desirable to enlarge the countersunk hole 56 and the bolt holes 58,60 to provide for more cooling or lower flow velocitites. The reactor vessel 10 may then be refueled and restarted. In alternative situations where the fluid flow passageways are formed by machining the baffle plate 40, the baffle plate 40 (and in some cases the former plate 42) may be machined after the existing bolts are removed, the same or similar bolts 50 installed and the reactor vessel 10 is then refueled. While various present preferred embodiments of the present invention has been shown and described, it is to be understood that the invention may be otherwise variously embodied within the scope of the following claims of invention.
	abstract	A method of mass producing nuclear fuel elements may include: forming a graphite base portion of the fuel elements; repeatedly performing a sequence of operations comprising depositing a uniform graphite layer over a previous layer, depositing a layer of particles on the uniform graphite layer within a fuel zone diameter, so that the particles are spaced apart in a predefined pattern, and applying a binder using additive manufacturing methods to bind each layer with successively increasing and then decreasing diameters to form a central portion of fuel elements including a fuel-containing fuel zone; and repeatedly performing a sequence of operations comprising forming a uniform graphite layer on a previous layer and applying a binder using additive manufacturing methods to bind each layer with successively decreasing diameters to form a cap portion of fuel elements. The particles may include one or more of a nuclear fuel material, burnable poison material, or breeder material. The fuel particles may be tri-structural-isotropic (TRISO) particles that do not have an overcoat.
047675721	abstract	Nuclear residues, such as concrete or metal parts of a nuclear reactor installation, are used as raw materials for the production of radiation shielding structures for such nuclear installation. The concrete residues can be broken up to form an aggregate for concrete which is cast to form such structures and metal objects can be added for the casting of transport and storage vessels for radioactive wastes.
	claims	1. An X-ray image detecting apparatus comprising: an X-ray grid arranged to remove X-rays scattered by a subject; a conversion member for converting X-rays passed by said X-ray grid into light having a predetermined wavelength; and a plurality of photoelectric conversion elements arranged to receive light from said conversion member, wherein said plurality of photoelectric conversion elements are arranged two-dimensionally spaced apart from each other and have predetermined insensitive regions disposed between said photoelectric conversion elements, said X-ray grid comprises a plurality of X-ray absorption members for removing scattered X-rays, each of said X-ray absorption members being arranged substantially only on said insensitive regions when viewed from a direction from which x-rays are incident, and said conversion member is arranged only in regions between said X-ray absorption members that are substantially adjacent to each other when viewed from the direction from which X-rays are incident. 2. The apparatus according to claim 1 , wherein said conversion member is partitioned by said X-ray absorption members. claim 1 3. The apparatus according to claim 1 , wherein said conversion member is partitioned by predetermined members which are different from said X-ray absorption members and located adjacent to each other. claim 1 4. The apparatus according to claim 3 , wherein said predetermined members have a property of substantially not transmitting the light from said conversion member. claim 3 5. The apparatus according to claim 1 , wherein said conversion member is partitioned by cavities, and wherein the surface of said conversion member facing said cavities has a property of substantially not transmitting the light from said X-ray conversion member. claim 1 6. The apparatus according to any one of claims 1 or 2 to 5 , wherein when viewed from a direction from which X-rays are incident, said X-ray absorption members are arranged in a stripe pattern and said conversion member is divided into parts, in a direction along said X-ray absorption members, in correspondence to respective ones of said photoelectric conversion elements. 7. The apparatus according to any one of claims 1 or 2 to 5 , wherein said X-ray grid is a converging grid having a predetermined focal point. 8. The apparatus according to any one of claims 1 or 2 to 5 , wherein said X-ray grid comprises intermediate substances disposed between said X-ray absorption members. 9. The apparatus according to claim 1 , further comprising: claim 1 a first member, for supporting one end of each of said X-ray absorption members on one surface of said first member as well as holding said conversion member on another surface of said first member; and a second member, for holding another end of each of said X-ray absorption members. 10. The apparatus according to claim 9 , wherein said first member separates and holds said conversion member so that said conversion member is disposed substantially only in regions between said X-ray absorption members that are adjacent to each other when viewed from a direction from which X-rays are incident. claim 9 11. The apparatus according to claim 10 , wherein said first member has a property of substantially not transmitting the light from said conversion member. claim 10 12. The apparatus according to claim 1 , wherein said photoelectric conversion elements are arranged on an insulation substrate. claim 1 13. The apparatus according to claim 1 , wherein said X-ray absorption members are arranged in a matrix or in a stripe pattern when viewed from a direction from which X-rays are incident. claim 1 14. The apparatus according to claim 1 , wherein said X-ray absorption members have a property of reflecting the light from said conversion member. claim 1 15. The apparatus according to claim 1 , wherein said X-ray grid has a grid ratio of at least 3:1. claim 1 16. An X-ray image acquisition apparatus comprising: an X-ray generator; and an X-ray image detector, wherein said X-ray image detector comprises: an X-ray grid arranged to remove X-rays scattered by a subject; a conversion member for converting X-rays passed by said X-ray grid into light having a predetermined wavelength; and a plurality of photoelectric conversion elements arranged to receive light produced by said conversion member, wherein said plurality of photoelectric conversion elements are arranged two-dimensionally with a predetermined insensitive region between each adjacent two of said photoelectric conversion elements, said X-ray grid comprising a plurality of X-ray absorption members for removing scattered X-rays, said X-ray absorption members are disposed substantially only on said insensitive regions when viewed from a direction from which X-rays are incident, and said conversion member is arranged only in regions between said X-ray absorption members that are substantially adjacent to each other when viewed from a direction from which X-rays are incident. 17. An X-ray image acquisition apparatus comprising: an X-ray image detector; and an image processor that receives and processes image data obtained using said X-ray image detector, wherein said X-ray image detector comprises: an X-ray grid; a conversion member for converting X-rays passed by said X-ray grid into light having a predetermined wavelength; and a plurality of photoelectric conversion elements arranged to receive light produced by said conversion member, wherein said plurality of photoelectric conversion elements are arranged two-dimensionally with a predetermined insensitive region between each adjacent two of said photoelectric conversion elements, said X-ray grid comprises a plurality of X-ray absorption members for removing scattered X-rays, said X-ray absorption members are disposed substantially only on said insensitive regions when viewed from a direction from which X-rays are incident, and said X-ray conversion member is arranged only in regions between said X-ray absorption members that are substantially adjacent to each other when viewed from a direction from which X-rays are incident.
	abstract	In a diagnostic X-ray system having a scattered radiation grid driven via a wobble bearing, restoring forces generated via the wobble bearing and hence the vibration on the diagnostic X-ray system are minimized while simultaneously requiring a minimal space. A plate-like compensatory mass disposed parallel to the scattered radiation grid is provided, which during operation executes a linear motion that is contrary to the linear motion of the scattered radiation grid.
054004993	summary	FIELD OF THE INVENTION This invention relates generally to maintenance of a control rod drive of a boiling water reactor. In particular, the invention relates to tools for dismantling or assembling a control rod drive during a maintenance operation. BACKGROUND OF THE INVENTION Control rod drives (CRDs) are used to position control rods in boiling water reactors (BWRs) to control the fission rate and fission density, and to provide adequate excess negative reactivity to shutdown the reactor from any normal operating or accident condition at the most reactive time in core life. Referring to FIG. 1, each CRD is mounted vertically in a CRD housing 10 which is welded to a stub tube 8, which in turn is welded to the bottom head of the reactor pressure vessel 4. The CRD flange 6 is bolted and sealed to the flange 10a of the CRD housing 10, which contains ports for attaching the CRD hydraulic system lines 80, 81. Demineralized water supplied by the CRD hydraulic system serves as the hydraulic fluid for CRD operation. As shown schematically in FIG. 1, the CRD is a double-acting, mechanically latched hydraulic cylinder. The CRD is capable of inserting or withdrawing a control rod (not shown) at a slow controlled rate for normal reactor operation and of providing rapid control rod insertion (scram) in the event of an emergency requiring rapid shutdown of the reactor. A locking mechanism in 12 the CRD permits the control rod to be positioned at 6-inch (152.4-mm) increments of stroke and to be held in these latched positions until the CRD is actuated for movement to a new position. A spud 46 at the top of the index tube 26 (the moving element) engages and locks into a socket at the bottom of the control rod. Once coupled, the CRD and control rod form an integral unit which must be manually uncoupled before a CRD or control rod may be removed from the reactor. When installed in the reactor, the CRD is wholly contained in housing 10. The CRD flange 6 contains an insert port 66, a withdraw port 70 and an integral two-way check valve (with a ball 20). For normal drive operation, drive water is supplied via an associated hydraulic control unit (HCU) to the insert port 66 for drive insertion and/or to withdraw port 70 for drive withdrawal. For rapid shutdown, reactor pressure is admitted to the two-way check valve from the annular space between the CRD and a thermal sleeve (not shown) through passages in the CRD flange, called scram vessel ports. The check valve directs reactor pressure or external hydraulic pressure to the underside of drive piston 24. Referring to FIG. 2, the CRD further comprises an inner cylinder 57 and an outer tube 56, which form an annulus through which water is applied to a collet piston 29b (see FIG. 1) to unlock index tube 26. The internal diameter of inner cylinder 57 is honed to provide the surface required for expanding seals 65 on the drive piston 24. Welded pipes 80 and 81, installed in the CRD housing, port water to the insert port 66 and the withdraw port 70 respectively. A port 69 below outer tube 56 connects to withdraw port 70 in CRD flange 6 so that water is applied through the annulus to collet piston 29b when a withdraw signal is given. The CRD is secured to the CRD housing flange 10a by eight mounting bolts (not shown). A pressure-tight seal is effected between the mated flanges by O-ring gaskets (not shown) mounted in a spacer 7 secured to the CRD flange face. Insert port 66 contains a ball check valve which consists of check-valve ball 20, ball retainer 21, and retainer O-ring 22. This valve directs HCU accumulator pressure or reactor pressure to the underside of drive piston 24 during scram operation. Port 66 is connected internally to the annulus and the bottom of drive piston 24 and serves as the inlet for water during normal insertion or scram. Water enters this port for a brief period in response to a withdraw signal to move the index tube 26 upward so that collet fingers 29a (see FIG. 1) are cammed out. Following this brief unlocking period, water from below drive piston 24 is discharged through port 66 and through the under-piston hydraulic line for the duration of the withdraw signal. The withdraw port 70 serves as the inlet port for water during control rod withdrawal and as the outlet port for water during normal or scram insertion. It connects with internal porting and annuli to the area above drive piston 24. During a withdraw operation, water is supplied from port 70 through a small connecting port in CRD flange 6 to the annular space between outer tube 56 and inner cylinder 57 for application to the bottom of collet piston 29b. The locking mechanism comprises collet fingers 29a, collet piston 29b and collet spring 31. This mechanism is the means by which index tube 26 is locked to hold the control rod at a selected position. The collet mechanism requires a hydraulic pressure greater than reactor pressure to unlock for CRD-withdraw movement. A preload is placed on collet spring 31 at assembly and must be overcome before the collet can be moved toward the unlocked position. For control rod withdrawal, a brief insert signal is applied to move index tube 26 upward to relieve the axial load on collet fingers 29a, camming them outward against the sloping lower surface of index tube locking notch 55. Immediately thereafter, withdraw pressure is applied. In addition to moving index tube 26 downward, this pressure is at the same time applied to the bottom of collet piston 29b to overcome the spring pressure and cam the fingers 29a outward against a guide cap (not shown). When the withdraw signal ceases, the spring pressure forces the collet downward so that fingers 29a slip off the guide cap. As index tube 26 settles downward, collet fingers 29a snap into the next higher notch and lock. When collet fingers 29a engage a locking notch 55, collet piston 29b transfers the control rod weight from index tube 26 to the outer tube 56. Unlocking is not required for CRD insertion. The collet fingers are cammed out of the locking notch as index tube 26 moves upward. The fingers 29a grip the outside wall of index tube 26 and snap into the next lower locking notch for single-notch insertion to hold index tube 26 in position. For scram insertion, index tube 26 moves continuously to its limit of travel during which the fingers snap into and cam out of each locking notch as index tube 26 moves upward. When the insert, withdraw or scram pressures are removed, index tube 26 settles back, from the limit of travel, and locks to hold the control rod in the required position. The drive piston 24 and index tube 26 are the primary subassembly in the CRD, providing the driving link with the control rod as well as the notches for the locking mechanism collet fingers. Drive piston 24 operates between positive end stops, with a hydraulic cushion provided at the upper end only. Index tube 26 is a nitrided stainless-steel tube threaded internally at both ends. The spud 46 is threaded to its upper end, while the head of the drive piston 24 is threaded to its lower end. Both connections are secured in place by means of bands 25 with tab locks. There are 25 notches machined into the wall of index tube 26, all but one of which are locking notches 55 spaced at 6-inch intervals. The uppermost surfaces of these notches engage collet fingers 29a, providing 24 increments at which a control rod may be positioned and preventing inadvertent withdrawal of the rod from the core. The lower surfaces of the locking notches slope gradually so that the collet fingers cam outward for control rod insertion. Drive piston 24 is provided with internal (62, 71, 72) and external seal rings (65), and is operated in the annular space between piston tube 15 and inner cylinder 57. Internal (63) and external (64) bushings prevent metal-to-metal contact between drive piston 24 and the surface of piston tube 15 and the wall of inner cylinder 57 respectively. The magnet housing, which comprises the lower end of drive piston 24, contains a ring magnet 67 which actuates the switches of the position indicator probe (not shown) to provide remote electrical signals indicating control rod position. The piston tube assembly forms the innermost cylindrical wall of the CRD. It is a welded unit consisting of piston tube 15 and a position indicator tube 61. The position indicator tube 61 is a pressure-containing part which forms a drywell housing for a position indicator probe 12a (see FIG. 2). Piston tube 15 provides for the porting of water to or from the upper end of the piston head portion of drive piston 24 during rod movement. The tube section 15a and head section 15b of piston tube 15 provide space for position indicator tube 61, which is welded to the inner diameter of the threaded end of head section 15b and extends upward through the length of tube section 15a, terminating in a watertight cap near the upper end of the tube section. Piston tube 15 is secured by a nut 16 at the lower end of the CRD. Two horizontal ports are provided in the head section 15b, 180.degree. apart, to transmit water between the withdraw porting in the CRD flange and the annulus between indicator tube 61 and tube section 15a of piston tube 15 for application to the top of drive piston 24. Three O-ring seals 18 are installed around head section 15b. Two seal the bottom of the CRD against water leakage and one seals the drive piston 24 under-piston pressure from the drive piston over-piston pressure. The position indicator probe 12a, which is slidably inserted into indicator tube 61, transmits electrical signals to provide remote indications of control rod position and CRD operating temperature. Probe 12a is welded to a plate 12b, which is in turn bolted to housing 12. Housing 12 is secured to the CRD ring flange 17 by screws 13. A cable clamp (not shown), located at the bottom of a receptacle 14, secures a connecting electrical cable to receptacle 14. Ring flange 17 is in turn secured to the CRD housing by screws 9. Thus, probe 12a, housing 12 and the cable clamp (with the cables passing therethrough) can be removed as a unit. In order to perform maintenance on a CRD, the CRD must be removed from its CRD housing. After the CRD and control rod are uncoupled, the CRD is removed from the CRD housing and then disassembled for the purpose of performing required maintenance on the respective CRD components. The CRD is placed in a horizontal position on a work table and disassembled. After drive piston 24 has been disengaged from the index tube and removed from the piston tube, it is disassembled into its component parts, i.e., magnet housing 24a, seal cups 24b, piston coupling 24c and piston head 24d (see FIG. 3). After piston coupling 24c has been disengaged from magnet housing 24a and piston head 24d, the internal bushings 63 must be removed from the piston coupling. These bushings are graphitar rings which sit in annular recesses located at opposing ends of the piston coupling bore. These rings become encrusted with crud during reactor operation. As a result, the internal bushings become stuck inside the piston coupling and are difficult to remove using conventional tools such as chisels and screwdrivers. Moreover, during such removal, the piston coupling may be damaged. SUMMARY OF THE INVENTION The present invention is a tool for removal of the internal bushings from a piston coupling of a CRD without causing damage to the piston coupling. The tool comprises a spring-loaded and hardened collet supported on one end of a ram. The collet has a pair of arms with shoulders or projections which latch inside the internal bushing ring to be removed. The tool is inserted into the piston coupling from either end, depending on which internal bushing is to be removed. The tool is inserted until the latching shoulders or projections snap behind the radially inwardly projecting internal bushing, with the contact surfaces of the shoulders or projections in contact with a radial end surface of the bushing. The bushing can then be dislodged, from the inside out, by impacting the other end of the ram with a hammer or mallet. The tool in accordance with the invention reduces the time required for CRD disassembly and consequently reduces the radiation exposure of maintenance personnel.
	description	In all the figures of the drawing, sub-features and integral parts that correspond to one another bear the same reference symbol in each case. Referring now to the figures of the drawing in detail and first, particularly, to FIGS. 1 and 2 thereof, there is shown a lance shaft 1 which is in the form of a tube. Measuring lances that carry sensors may be disposed in the lance shaft 1. The sensors determine the state inside a reactor pressure vessel. The lance shaft 1 is inserted into the reactor pressure vessel from outside in a pressure-tight manner. To this end, it is guided in a nozzle 2 of a cover of the reactor pressure vessel and is sealed off from the nozzle 2. The sealing is ensured by pressing a sealing ring 5 via an integrally formed portion 4 on the lance shaft 1 into a seat 10 on a threaded socket 14 connected to the nozzle 2, in the course of which the sealing ring 5 is likewise pressed against the nozzle 2. For pressing the sealing ring 5 into place, a force is exerted on a pressing plate 3 which bears against the nozzle 2, a factor which leads to an opposed force being exerted on the lance shaft 1 by a mechanical force transmission, and this force presses the integrally formed portion 4 of the lance shaft 1 against the seat 10. In the known device (FIG. 1), the force on the pressing plate 3 is produced via pressing screws 11. For the mechanical force transmission to the lance shaft 1 there is a plate 15, which is firmly connected to the lance shaft 1 and has threads for accommodating the pressing screws 11. When the pressing screws 11 are screwed in, the plate 15 and thus also the lance shaft 1 are lifted. The combination of the pressing screw 11 and the plate 15 constitutes the basic form of a pressure element. The pressing screws 11 do not act directly on the pressing plate 3 but on an additional pressing plate 12, one or more disk springs 6 being located between the pressing plate 3 and the additional pressing plate 12. The disk springs 6 ensure a preloading force, which is supplemented by the force of the pressing screws 11. The pressure element may consist of the pressing screw 11, the plate 15 and the disk springs 6. The preloading force of the disk springs 6 must be reset after every inspection, a factor which necessitates additional work on the device, and this additional work must be carried out with protective precautions taken. The device according to the invention (FIG. 2) correspondingly has the additional pressing plate 12 and one or more of the disk springs 6 which are disposed in the same way. The difference from the known device consists in the fact that pressure parts 7 that are hydraulically actuated take the place of the pressing screws 11. The pressure parts 7 may be, for example, so-called HYTORC nuts. The pressure part 7 alone constitutes the basic form of a pressure element. The pressure part 7 and the disk springs 6 may then together form the pressure element, the pressure part 7 producing a force and the disk springs 6 producing a preloading force (preloading). For the mechanical force transmission from the pressure part 7 to the lance shaft 1, the pressure part 7 has two parts moving in opposite directions. If the outer part 7a is moved downward, the inner part 7b is lifted and takes the lance shaft 1 with it, since it is firmly connected to it. For driving the pressure parts 7 there is provided a hydraulic tool 16 (FIG. 3), which is connected to a hydraulic line 8, in which a pump 13 for the hydraulic medium is fitted. A pressure gauge 9 is connected to the hydraulic line 8. The pressure gauge 9 measures the pressure of the hydraulic medium, which enables the instantaneous preloading force of the disk springs 6 to be inferred. The preloading force of the disk springs 6 can be set by varying the pressure by the pump 13. The hydraulic tool 16 covers or surrounds the top part of the lance shaft 1 and the entire pressure part 7 and rests on the additional pressing plate 12, which, configured as a ring, surrounds the lance shaft 1 shown in FIG. 2. The lance shaft 1 extends inside the tubular nozzle 2, a factor which cannot be seen in FIG. 3, and if need be also inside the pressing plate 3, the disk springs 6 and the additional pressing plate 12, all of which surround the lance shaft 1 in an annular manner. The hydraulic tool 16 (FIG. 3) may have an opening through which the lance shaft 1 can pass upward. With the device according to the invention, on the one hand sealing of the lance shaft 1 in the nozzle 2 can be effected quickly and reliably in a remote-controlled manner solely via the hydraulic line 8 and with the pump 13, and in addition the preloading force of the disk springs 6 can be monitored by the pressure gauge 9 and can be varied or set by the pump 13. Advantageously, it is not necessary for personnel in protective suits to work directly at the leadthrough.
	summary
062122509	abstract	A method for providing a leak-tight metal enclosure to a fuel matrix penetrated by coolant channels, wherein the mutually contacting surfaces of said metal enclosure and said fuel matrix are metallurgically bonded, comprising placing a metal cladding about the lateral surface of said fuel matrix; disposing metal coolant tubes within said coolant channels; placing a perforated header plate having tubular extensions at each end of the fuel matrix from which the coolant tube ends protrude, said coolant tubes passing through said perforated header plate and said tubular extensions and terminating even with the ends of said extensions; welding, under vacuum, said cladding to said header plates, and the ends of said coolant tubes to the ends of said tubular extensions; exposing the assembly comprising the fuel matrix and enclosure to a gas at high temperature and pressure; and machining said header plates to provide a finished fuel element.
039393534	summary	BACKGROUND OF THE INVENTION The present invention relates to an apparatus for mounting a specimen in an electron microscope. Generally, with known specimen mounting apparatus of this type, a specimen chamber is evacuated while the electron microscope is in use, and is returned to atmospheric pressure when the specimen is changed. This necessitates that the position of the specimen be adjusted from outside the chamber. This has resulted in apparatus that is complicated in construction, troublesome to handle, and sensitive to vibrations which cause a focusing problem. In addition conventional specimen mounting apparatus need complicated structure to indicate the position of the electron beam of an electron microscope on a specimen during use of the electron microscope. SUMMARY OF THE INVENTION The object of the present invention is to solve the aforementioned problems in conventional specimen mounting apparatus by providing a specimen mounting apparatus capable of simple construction and ease of handling. Another object of the present invention is to provide a specimen mounting apparatus which is releasably coupled to the electron lens of an electron microscope to prevent the relative movement of one with respect to the other to ensure proper focusing when the specimen mounting apparatus is induced to vibrate. A further object of the present invention is to provide a specimen mounting apparatus which can easily adjustably position a specimen to be irradiated by an electron beam to select a portion of the specimen to be irradiated. A still further object of the present invention is to provide a specimen mounting apparatus which is capable of indicating the relative position of the specimen to be irradiated to thereby indicate the portion of the specimen that has been selected to be irradiated. Still another object of the present invention is to provide a specimen mounting apparatus which can be easily inserted and removed to change the specimen. These and the other objects of the present invention are carried out by the specimen mounting apparatus of the present invention which comprises specimen-holding means positionable into an irradiating position wherein a portion of the specimen is exposed to the electron beam. The specimen-holding means includes means for adjustably positioning the specimen, when the specimen-holding means is in the irradiating position, to variably select the portion of the specimen exposed to the electron beam. The apparatus also includes means for releasably mounting the specimen-holding means in the irradiating position and for releasably coupling the specimen-holding means to the electron lens, when the specimen-holding means is in the irradiating position, to transmit vibratory motion from the specimen-holding means to the lens to substantially prevent the relative movement of one with respect to the other whenever the specimen-holding means is induced to vibrate, thus maintaining the focus of the electron beam on the selected portion of the specimen. The specimen-holding means comprises a rotatable, elongated, cylindrical member having means disposed at one end portion thereof for mounting the specimen. The elongated member is fluidtightly mounted for rotation about its longitudional axis in a rotatable cylindrical support member which has a main axis about which it is mounted to rotate. The elongated member is inserted in the support member in a throughbore that has an axis parallel to the main axis of the support member and disposed eccentrically thereof. The means for coupling the specimen-holding means to the electron lens comprises means for forcibly abutting a portion of the specimen-holding means on the electron lens. In addition, an intermediate member which contacts the electron lens may be provided and in this case means are provided for forcibly abutting a portion of the specimen-holding means on the intermediate member. The means for releasably mounting the specimen-holding means comprises an evacuatable specimen chamber that is receptive of a negative pressure applied thereto and has means therein defining a throughbore for fluidtightly receiving the specimen-holding means. Thus the specimen-holding means is forcibly held in the wall of the specimen chamber when a pressure differential is developed between the chamber and the atmosphere by evacuating the specimen chamber. The apparatus further comprises indicating means for indicating the relative position of the specimen thereby indicating the portion of the specimen exposed to the electron beam. The indicating means comprises a pilot member with a pilot area approximately as large as the means for mounting the specimen that is disposed at the one end portion of the elongated member. The pilot member is attached to the other end portion of the elongated member. An indicating member is provided with an indicating part that moves in correspondence with the portion of the specimen relative to the electron beam. The indicating part is located close to the surface of the pilot area. Having in mind the above and other objects that will be obvious from an understanding of the disclosure, the present invention comprises the combinations and arrangements of parts illustrated in the presently preferred embodiments of the present invention which is here and after set forth in sufficient detail to enable those persons skilled in the art to clearly understand the function, operation, construction, and advantages of it when read in conjunction with the accompanying drawings.
051757559	summary	BACKGROUND OF THE INVENTION X-ray lithography utilizes a variety of sources including X-rays emitted from a small area (point-sources) and synchrotron generated X-rays to generate an image. Unfortunately, X-ray lithographic systems have been limited by the inability to adequately manipulate the X-ray beam. X-ray optics incur several difficulties not encountered in the visible or infra-red (IR range). Refraction in passing through media of a different refractive index cannot be used because of the strong absorption of photons with sufficient energy to excited or ionize electronic levels inside the media. Diffraction and interference phenomena can be used to deflect X-rays using Bragg scattering in single crystals, in multi-layer mirrors or by using zone and phase plates. Although these approaches are useful in many applications, they are very energy (wave length) selective and cannot be used to control X-ray beams having a broad energy spectrum. The use of reflection has also been limited because surfaces of all known materials have very low reflection coefficients for X-radiation at large angles of incidence. Grazing-incidence optics have been developed based on the phenomenon of total external reflection of X-rays. This is widely used in synchrotron radiation facilities where flat mirrors are used for deflection and curved mirrors are used for focusing parallel beams. These mirrors typically use a single reflection. Such devices have an extremely small angular aperture due to the small value of the total-external-reflection angle (milliradians at KeV energies). Point-source X-ray lithography using existing equipment is limited by the following: Intensity. The sources currently in development lack the intensity to achieve an exposure time which approaches production requirements. Modifications attempted to increase intensity are not only expensive, but by pushing the sources harder potentially decrease reliability, reduce source life and increase debris generation at the source which can damage the mask. PA1 Radial magnification. Because the beam from the source to the mask is divergent there is increasing distortion as the edge of the field is reached. Distortion may be reduced by adjusting the feature size and shape in the mask. Unfortunately, as gap tolerance becomes more critical, mask and wafer flatness requirements increase, alignment becomes increasingly difficult, field size is limited, and the same masks cannot be used on a synchrotron. PA1 Penumbral blur. The sources have a size large enough that illumination of the mask by different points of the x-ray generating area produce a blurring of the features projected on to the wafer. This lack of definition in the edge of the images projected limits the achievable minimum feature size. PA1 Source position instability. To the extent that X-ray spots are not be in the exact same position each pulse, feature patterns projected on the resist have decreased definition. Synchrotron-source X-ray lithography is not intensity limited and has a beam which does not show significant divergence of any significance in the vertical direction. The beam, however, is very flat, normally 0.5-2.0 mm thick, is horizontal, and has a divergence in the horizontal plane which can be 6 degrees or larger. Because the beam is flat and the area to be exposed can be multiples of a cm square, either the wafer and mask must be moved to get a scan or a mirror in the beam line must be rotated to cause the beam to scan across the desired area. Horizontal beams require that the masks and wafer be vertical rather than horizontal as is more commonly used with optical steppers. The horizontal beam divergence causes the majority of the beam to be wasted with only a small portion of the beam reaching the mask and wafer at the end of the long beam lines. The subject invention provides a solution to the long felt need in the art for an improved system of X-ray lithography. The subject invention provide the benefits of improved X-ray control, precision and accuracy. SUMMARY OF THE INVENTION The subject invention provides an X-ray lithographic system comprising a Kumakhov lens. An X-ray source is required and the Kumakhov lens is typically located between X-ray source and a mask. The X-ray source may be a point source or a non-point source, such as a synchrotron. A Kumakhov lens may also be located between a mask and a resist. The subject invention also teaches a method for X-ray lithography, which comprises: providing a source of radiation; focusing the radiation from the source through a Kumakhov lens; and passing the focused radiation through a mask. This method may also add a Kumakhov lens to form a quasi-parallel beam and a second Kumakhov lens to focus the beam into a preselected band of energies.
	description	This specification relates to controlling the temperature of a uranium material in a uranium enrichment cycle and particularly, but not exclusively, to controlling the temperature of uranium hexafluoride (UF6) inside industry-standardized 48Y and 30B UF6 cylinders in a uranium enrichment facility. In a uranium enrichment facility, uranium material is heated and cooled in industry-standardized transport cylinders before and after being fed through the enrichment apparatus. This specification provides an apparatus arranged to control the temperature of uranium material in a uranium material storage container, comprising a thermal guide which wraps around an external surface of the uranium material storage container to cause a heat transfer medium inside the thermal guide to exchange heat energy with the uranium material storage container; and a heat exchanger to heat or cool the heat transfer medium outside the thermal guide. The thermal guide may form a thermally conductive contact with the uranium material container to cause the exchange of heat energy by conduction. The exchange of heat energy may increase or decrease the temperature of the uranium material. The thermal guide may be configured to guide the heat transfer medium around the exterior of the uranium material container. The apparatus may be configured to cause the heat transfer medium to flow between the thermal guide and the heat exchanger. The thermal guide may surround the uranium material container. The thermal guide may comprise a thermally conductive heat transfer surface for locating against the external surface of the uranium material container and through which heat energy is exchanged between the heat transfer medium in the guide and the uranium material container. The thermal guide may comprise a heat insulating surface which is configured to prevent heat transfer between the heat transfer medium in the guide and the atmosphere around the uranium material container. The apparatus may be configured to controllably heat or cool the heat transfer medium in order to cause heating or cooling of the uranium material inside the uranium material container. The apparatus may be configured to detect the temperature of the uranium material container and to heat or cool the heat transfer medium in response to the detected value of the temperature of the uranium material container. The apparatus may be configured to heat or cool the heat transfer medium to obtain a predetermined target temperature for the uranium material container. The target temperature may be sufficient to cause the uranium material inside the uranium material container to change material state. The apparatus may be arranged to circulate the heat transfer medium between the thermal guide and the heat exchanger to heat or cool the heat transfer medium. The thermal guide may be selectively attachable to, and/or releasable from, the exterior of the uranium material container. The apparatus may comprise quick-release connections which allow the thermal guide to be selectively attached to, and/or released from, the uranium material container. The quick-release connections may comprise magnetic connections which attach the thermal guide to the uranium material container. The quick-release connections may comprise mechanical clamps which attach the thermal guide to the uranium material container. The thermal guide may comprise a plurality of sections which wrap around a corresponding plurality of regions of the uranium material container. The heat transfer medium may be a liquid. The heat transfer medium may be a gas. This specification also provides a uranium material storage container wrapped in the thermal guide. The apparatus may comprise a further thermal guide which wraps around an external surface of a further uranium material storage container to cause a heat transfer medium inside the further thermal guide to exchange heat energy with the further uranium material storage container, wherein the heat exchanger is configured to transfer heat energy extracted from the warmer storage container to the cooler storage container. This specification also provides a method of controlling the temperature of uranium material in a uranium material storage container, comprising wrapping the uranium material storage container in a thermal guide; and heating or cooling a heat transfer medium outside the thermal guide to cause the heat transfer medium to exchange heat energy with the uranium material storage container when inside the guide. A thermal energy transfer apparatus 1 for safely heating and cooling uranium hexafluoride (UF6) in a uranium enrichment facility is described below. The apparatus 1 is adapted to heat and cool the UF6 inside industry-standardized uranium material containers 2 that have been manufactured and certified according to ISO and ANSI specifications. The examples below discuss the containers 2 in the context of 48Y UF6 cylinders 2 and 30B UF6 cylinders 2. Both of these types of cylinder 2 can be used to contain UF6 at depleted, natural or enriched concentrations of U235. When industry-standardized UF6 containers 2 are exposed to normal atmospheric temperatures, for example during long term storage and transportation, the conditions inside the containers 2 are such that the UF6 is in a solid state, with a vapour pressure of approximately 100 mbar. The thermal energy transfer apparatus 1 described herein is arranged to convert UF6 inside the containers 2 from a solid state to a gaseous state when being fed from the containers 2 into a uranium enrichment apparatus, such as a cascade of gas centrifuges. The thermal energy transfer apparatus 1 is also arranged to convert enriched and depleted UF6 products of the enrichment apparatus from a gaseous state back into a solid state inside the containers 2. As explained in detail below, the thermal energy transfer apparatus 1 effects the conversions in the state of the UF6 by conducting heat energy into and out of the walls of the containers 2 in a safe and energy efficient manner. Referring to FIGS. 1 and 2, an industry-standardized UF6 container 2 is approximately cylindrical in shape and comprises a longitudinal wall 3 and two end walls 4, 5. The end walls 4, 5 are located at opposite ends of the container 2 and the cylindrical longitudinal wall 3 extends between them. The perimeter 6 of each end wall 4, 5 is approximately circular and is joined to the cylindrical longitudinal wall 3. The exteriors of the end walls 4, 5 form the exterior end surfaces 7 of the container 2, and the exterior of the longitudinal wall 3 forms the exterior longitudinal surface 8 of the container 2. In FIG. 1, the external diameter of the container 2 is illustrated as being approximately constant along the container's length. However, the skilled person will be aware that in some types of standardized UF6 container 2 the diameter of the container 2 narrows towards either end. The walls 3, 4, 5 of the container 2 are thermally conductive and thus allow heat energy to be transferred into and out of the UF6 material through the walls 3, 4, 5. The longitudinal wall 3 includes a plurality of circumferential stiffening ribs 9 which extend around the cylinder 2 at regular intervals along its length. The orientation of the ribs 9 is approximately parallel to the end walls 4, 5 of the cylinder 2, such that the ribs 9 project outwardly in a direction that is approximately perpendicular to the surface 8 of the longitudinal wall 3. The ribs 9 divide the exterior surface 8 of the longitudinal wall 3 into a plurality of cylindrical sections 8A-D. The boundaries of each section 8A-D are defined either by a pair of ribs 9 or by a rib 9 on one side and an end of the cylinder 2 on the other side. The cylindrical wall sections 8A-D each extend fully around the circumference of the UF6 cylinder 2 and may be approximately equal in length. For example, the UF6 cylinder 2 illustrated in FIG. 1 comprises three circumferential ribs 9 that, together, divide the longitudinal exterior surface 8 of the cylinder 2 into four cylindrical sections 8A-D. FIGS. 3 and 4 illustrate an industry-standardized 48Y UF6 cylinder 2 in thermal contact with the thermal energy transfer apparatus 1. More specifically, in FIGS. 3 and 4, the exterior surfaces 7, 8 of the UF6 cylinder 2 are in contact with a thermal guide 10 of the thermal energy transfer apparatus 1. The thermal guide 10 is flexible in shape and is wrapped around the UF6 cylinder 2 so that the exterior surfaces 7, 8 of the UF6 cylinder 2 are encompassed by the guide 10. As explained below, the guide 10 contains a heat transfer medium 11 which exchanges heat energy with the UF6 through the walls 3, 4, 5 of the UF6 cylinder 2 to change the material state of the UF6 pre and post enrichment. Referring to FIG. 5, the guide 10 comprises a heat transfer surface 12, a heat transfer medium containing region 13 and a heat insulating surface 14. The heat transfer medium containing region 13 is located in the interior of the guide 10, between the heat transfer surface 12 and the heat insulating surface 14. Both surfaces 12, 14 of the guide 10 are impermeable to the heat transfer medium 11. This prevents contact between the heat transfer medium 11 and the exterior surfaces 7, 8 of the UF6 cylinder 2. It also prevents contact between the heat transfer medium 11 and the atmospheric air around the outside of the guide 10. The heat transfer surface 12 is located against the external surfaces 7, 8 of the UF6 cylinder 2 and is thermally conductive. It may, for example, comprise the surface of a flexible, thermally permeable membrane 15 at the exterior of the guide 10. The thermally permeable membrane 15 may be elastic in order to ensure a consistent thermally conductive contact with the exterior surfaces 7, 8 of the UF6 cylinder 2. The thermally conductive contact between the heat transfer surface 12 and the exterior surfaces 7, 8 of the UF6 cylinder 2 causes heat energy to conduct through the heat transfer surface 12 between the heat transfer medium 11 and the exterior walls 3, 4, 5 of the cylinder 2. The rate and direction of the heat conduction is dependent on the temperature gradient between the heat transfer medium 11 in the guide 10 and the external surfaces 7, 8 of the UF6 cylinder 2. Therefore, as described in more detail below, the rate and direction of thermal energy transfer between the UF6 in the cylinder 2 and the heat transfer medium 11 in the guide 10 can be controlled by controlling the temperature of the heat transfer medium 11. The heat transfer surface 12 follows the external contours of the cylinder 2 so that the nature of its contact with the exterior surfaces 7, 8A-D is continuous and encompassing. For example, as illustrated in FIGS. 3 and 4, the thermal guide 10 may extend around the full circumference of the cylinder 2 so that the heat transfer surface 12 is in contact with the exterior surface 8 of the longitudinal wall 3 around the full circumference of the wall 3. The length and width of the guide 10 are matched specifically with the corresponding dimensions of the cylinder 2 so as to provide an uninterrupted contact with the external surfaces 7, 8 of the cylinder 2. The thickness of the guide 10, i.e. the distance between the heat transfer surface 12 and the heating insulating surface 14, may be between approximately 1 cm and approximately 5 cm, such as between approximately 2 cm and approximately 3 cm. The continuous contact between the external surfaces 7, 8 of the cylinder 2 and the heat transfer surface 12 allows thermal energy to be conducted between the heat transfer medium 11 and the UF6 cylinder 2 over a high proportion of the total external surface 7, 8 of the cylinder 2. The conductive nature of the thermal exchange and the encompassing nature of the guide 10 around the cylinder 2 may provide for a high degree of efficiency in the thermal energy transfer and thus lower the amount of energy required for the UF6 to be cooled or heated, as desired. The conductive thermal exchange and encompassing nature of the guide 10 may also allow for the temperature of the cylinder 2 to be changed rapidly and thus controlled with a high degree of accuracy. The heat insulating surface 14 is located on the opposite side of the guide 10 to the heat transfer surface 12 so that it faces outwards from the cylinder 2. The heat insulating surface 14 is not thermally conductive and therefore substantially prevents heat energy from being exchanged between the air around the outside of the guide-wrapped cylinder 2 and the heat transfer medium 11 in the guide 10. The thermally insulating nature of the insulating surface 14 may further increase the efficiency of the thermal energy transfer between the heat transfer medium 11 and the UF6 cylinder 2. The flexible nature of the thermal guide 10 allows it to be added to the UF6 cylinder 2 by wrapping it around the exterior surfaces 7, 8 of the cylinder 2. Similarly, the flexible nature of the thermal guide 10 allows it to be removed from the UF6 cylinder 2 by unwrapping it from the exterior surfaces 7, 8 of the cylinder 2. In this way, the thermal guide 10 can be selectively attached to, and released from, the UF6 cylinder 2. The addition and removal of the guide 10 to and from the cylinder 2 can be rapidly achieved because the guide 10 is connected to the cylinder 2 using quickly attachable and releasable connectors 16, as illustrated in FIG. 6. These connectors 16 may, for example, secure the guide 11 directly to sections of the exterior surfaces 7, 8 of the cylinder 2. Additionally or alternatively, as illustrated in FIG. 6, the connectors 16 may secure the guide 10 to the ribs 9. This is convenient because it avoids any disruption that could be caused by the connectors 16 to the thermally conductive contact between the heat transfer surface 12 and the longitudinal exterior surfaces 7, 8 of the UF6 cylinder 2. The connectors 16 may be magnetic connectors 16. For example, the guide 10 may comprise magnetic regions 16 which magnetically adhere to the carbon-steel material of a 48Y or 30B UF6 cylinder 2. Alternatively, the connectors 16 may comprise another type of releasable fixing such as releasable clamps. In some embodiments, for example when the UF6 cylinder 2 comprises the ribs 9 shown in FIG. 1, the guide 10 comprises a plurality of separate longitudinal sections 10A-D. These longitudinal sections 10A-D comprise a plurality of separate lengths of the guide 10 that are respectively wrapped around different cylindrical sections 8A-D of the longitudinal surface 8 of the cylinder 2. An example of this is illustrated in FIG. 3. The dimensions of the guide sections 10A-D match those of the cylindrical sections 8A-D of the cylinder 2 that they are intended to cover so that only the ribs 9 of the cylinder 2 remain exposed. The guide 10 may additionally or alternatively comprise two separate end sections 10E-F, which respectively cover the end surfaces 7 of the cylinder 2. An example of this is illustrated in FIG. 4. As with the separate longitudinal sections 10A-D described above, and the guide 10 generally, the dimensions of the end sections 10E-F of the guide 10 match those of the surfaces 7 of the cylinder 2 that they are intended to cover. In this way, the guide 10 covers substantially the complete external surface 7, 8 of the cylinder 2. The guide sections 10A-F can each be added to and removed from the cylinder 2 separately from one another using the magnetic connections 16 referred to above. The guide 10 is re-usable and so, in the uranium enrichment facility, the guide 10 can be used to heat or cool a plurality of UF6 cylinders 2 in sequential order. A plurality of the guides 10 can thus be used to provide a consistent supply of heated UF6 material for enrichment and a correspondingly consistent cooling of UF6 material received from the enrichment apparatus post enrichment. For example, once a particular one of the guides 10 has been used to heat the UF6 material in a particular (e.g. 48Y) cylinder 2 to the desired temperature for use in the next stage of the uranium enrichment process, the guide 10 can be removed from the cylinder 2 by releasing the connections 16 referred to above and unwrapping it from the cylinder's surface 7, 8. The guide 10 can then be attached to another (e.g. 48Y) cylinder 2 in order to heat the UF6 inside the new cylinder 2 in the same manner as the previous cylinder 2. The process may be repeated as often as is necessary to provide the desired rate of gaseous UF6 for use in the next stage of the enrichment cycle. Similarly, once a guide 10 has been used to cool a (e.g. 30B) cylinder 2 of post enrichment UF6 to the desired temperature, causing the UF6 to convert from a gaseous state back to a solid state, the guide 10 can be removed from the cylinder 2 and attached to another (e.g. 30B) cylinder 2 to cool another quantity of post enrichment UF6 in the same manner. The weight of the guide 10 is such that it can be attached to and removed from the UF6 cylinders 2 by a human operator. For example, the mass of each section 10A-F of the guide 10 may be between approximately 5 kg and approximately 20 kg, such as between approximately 10 kg and approximately 15 kg. The heat transfer medium 11 is a fluid in either liquid or gaseous form. For example, the heat transfer medium 11 may be air or a medium with a higher heat capacity such as water or glycol. As described below, the thermal energy transfer apparatus 1 is configured to control the temperature of the heat transfer fluid 11 in order to control the flow of heat energy through the heat transfer surface 12 and thereby to accurately control the temperature of the UF6 inside the UF6 cylinder 2. Referring to FIG. 7, the temperature of the heat transfer fluid 11 is controlled by causing the heat transfer fluid 11 to continuously flow through a looped heat exchange path 17. The looped path 17 comprises a fluid channel circuit, which includes the heat transfer medium containing region 13 in the guide 10 and a heat exchanger 18 outside the guide 10. The heat exchanger 18 may, for example, be located in the hall which houses the UF6 take-off and/or feed-stations for the enrichment apparatus. The heat exchanger 18 may be configured to draw heat energy from, and/or expel heat energy to, the external atmosphere around the heat exchanger 18, such as that in or outside the hall, in order to heat or cool the heat transfer fluid 11 as required. The heat exchanger 18 may, for example, comprise a heat pump 18. The heat transfer fluid 11 is continuously directed around the circuit from the heat exchanger 18 to the containing region 13 of the guide 10 and then back to the heat exchanger 18. A suitable fluid pump (not shown) may be used to circulate the heat transfer fluid 11. Referring to FIG. 8, the heat exchanger 18 may be coupled to fluid channel circuits 17 of both a UF6 feed station, in which one or more UF6 cylinders 2 are heated to feed gaseous UF6 to the enrichment apparatus, and a UF6 take-off station, in which one or more UF6 cylinders 2 are cooled to solidify gaseous UF6 taken-off from the enrichment apparatus. For example, the heat exchanger 18 may be configured to extract heat energy from heat transfer fluid 11 in the fluid channel circuit 17 of the UF6 take-off station and to add heat energy to heat transfer fluid 11 in the fluid channel circuit 17 of the UF6 feed station. The heat exchanger 18 may be configured to transfer the heat energy that is extracted from the heat transfer fluid 11 in the circuit 17 of the UF6 take-off station into the heat transfer fluid 11 in the circuit 17 of the UF6 feed station. In this way, the heat transfer fluid 11 in the take-off station circuit 17 is cooled at the heat exchanger 18 in order to cause the fluid 11 to cool UF6 cylinders 2 in the take-off station. Conversely, the heat transfer fluid 11 in the feed station circuit 17 is heated at the heat exchanger 18 in order to cause the fluid 11 to heat UF6 cylinders 2 in the feed station. The thermal energy used to heat the heat transfer fluid 11 in the feed station circuit 17 is thereby at least partially drawn from the high temperature UF6 being received at the take-off station from the enrichment apparatus. The extraction of heat energy from the high energy UF6 in the take-off station for use in heating the low energy UF6 in the feed-station makes the heating and cooling process both energy efficient and environmentally advantageous because the heat energy extracted from the UF6 in the take-off station is not wastefully expelled to the open atmosphere. It will be appreciated that the exchange of heat energy in the heat exchanger 18 may be used to maintain, rather than to substantially increase or decrease, the temperatures of the UF6 cylinders 2 in the UF6 take-off and feed stations and/or the temperature of the heat transfer fluid 11 in the fluid circuits 17. The transfer of heat energy between one or more UF6 cylinders 2 in one or more feed stations and one or more UF6 cylinders 2 in one or more take-off stations, as described above, may be used to achieve such a temperature maintenance effect. Referring to FIG. 9, the heat transfer medium containing region 13 of the guide 10 may comprise one or more fluid channels 13A in thermally conductive contact with the thermally permeable membrane 15 located against the external surfaces 7, 8 of the cylinder 2. For example, in operation, the heat transfer fluid 11 may be piped along a circulation line 19 from the heat exchanger 18 into the guide 10 and divided amongst a plurality of heat transfer tubes 13A that together direct the heat transfer fluid 11 to all regions of the guide 10 before it is piped back along the circulation line 19 to the heat exchanger 18. The even distribution of the tubes 13A in the guide 10 provides a correspondingly even level of heat exchange over the external surface area 7, 8 of the UF6 cylinder 2. In the case where the thermal guide 10 comprises a plurality of individual sections 10A-F of the type described above, each of the sections 10A-F may comprise a plurality of such fluid channels 13A. Alternatively, the heat transfer medium containing region 13 may comprise a cavity which is bounded by the walls of the thermal guide 10. The cavity may be substantially uninterrupted across the area of the guide 10 so that the heat transfer fluid 11 piped into the cavity via the circulation line 19 fills the cavity and causes heat exchange to take place evenly over the surfaces 7, 8 of the cylinder 2. If the guide 10 comprises a plurality of sections 10A-F, as described above, then each section 10A-F may comprise its own cavity which is individually filled by fluid 11 piped from the heat exchanger 18. As illustrated in FIG. 7, the heat exchanger 18 is communicatively coupled to a controller 20, such as an electronic microcontroller 20, which is configured to control the operation of the heat exchanger 18. In particular, the controller 20 is configured to control the rate and direction of the flow of heat energy into or out of the heat transfer fluid 11 in the heat exchanger 18 in order to control the temperature of the fluid 11 and, in doing so, to control the temperature of the UF6 material inside the cylinder 2. In order to do this, the controller 20 may store in a memory 20A a target temperature for the interior of the UF6 cylinder 2 and cause the heat exchanger 18 to transfer heat energy into and/or out of the heat transfer fluid 11 in order to obtain and/or maintain the target temperature inside the UF6 cylinder 2. The controller 20 may continuously or regularly monitor the temperature of the cylinder 2 using one or more temperature sensors 21 on the cylinder 2. The temperature sensors 21 are communicatively coupled to the controller 20 to communicate temperature measurements to the controller 20. The controller 20 uses the temperature measurements from the sensors 21 to vary the operation of the heat exchanger 18 in order to achieve an appropriate rate of heating or cooling. For example, if the temperature sensed by the sensors 21 is below the target temperature for the UF6 cylinder 2, the controller 20 may cause the heat exchanger 18 to direct more heat energy into the heating fluid 11 to increase its temperature. Likewise, if the temperature inside the cylinder 2 is sensed by the sensors 21 to be above the target temperature, the controller 20 may cause the heat exchanger 18 to remove heat energy from the heating fluid 11 to decrease its temperature. The cylinder 2 may also comprise one or more pressure sensors 22 that are configured to determine the internal pressure of the cylinder 2 and are communicatively coupled to the controller 20 to communicate pressure measurements to the controller 20. The controller 20 uses the pressure measurements to monitor the internal pressure of the cylinder 2 to ensure that it correlates with an expected pressure value stored in the memory 20A. For example, the controller 20 may use the pressure measurements to ensure that the pressure of the cylinder 2 is in the region of 400 mbar. The target temperature stored at the controller 20 for the UF6 cylinder 2 is set so as to cause the UF6 inside the cylinder 2 to change state between gas and solid as required. For example, during heating of the UF6 material pre-enrichment, the controller 20 may be configured to cause the UF6 material to be heated to a temperature of between 40° C. and 60° C., such as approximately 55° C., in order to cause the UF6 inside the cylinder 2 to change from solid to gas inside the cylinder 2. If, as intended, the thermal energy transfer apparatus 1 is used in open environments where the UF6 cylinder 2 is not contained in a sealed system, the controller 20 is configured to limit the temperature of the UF6 to values below its triple point temperature of 64° C. for safety reasons. For example, the controller 20 and heat exchanger 18 may be configured to ensure that the temperature of the heat transfer fluid 11 also remains below 64° C. by implementing a temperature-based cut-off in the heat exchanger 18. During cooling of the UF6 material post enrichment, the controller 20 may be configured to cause the UF6 material to be cooled to a temperature below 40° C. An example temperature is between 20° C. and −25° C., although the apparatus 1 could be used to cool the UF6 to lower temperatures if desired. The target temperature is user controllable and can be set by inputting a command to the controller 20 via a user interface 23 of the thermal energy transfer apparatus 1. For example, the apparatus 1 may comprise a control panel 23 through which the commands can be entered. In addition to the temperature of the UF6 cylinder 2, the controller 20 may also monitor the temperature of the heat transfer fluid 11 directly in order to allow it to effect accurate temperature adjustments to the fluid 11 at the heat exchanger 18. In this way, the controller 18 can make correspondingly accurate adjustments to the temperature of the UF6 cylinder 2, for example based on a relationship between the temperature of the fluid 11 and the temperature of the cylinder 2 which is stored in the memory 20A. The controller 20 may monitor the temperature of the fluid 11 using temperature sensors (not shown) located in the looped heat exchange path 17. Such sensors may be located, for example, in the heating medium containing region 13 of the thermal guide 10, in the heat exchanger 18 and/or in the fluid circulation line 19. The controller 20 may be comprised within a Plant Control System which, in addition to monitoring and controlling the temperature and pressure of the UF6 cylinders 2 as referred to above, is additionally configured to monitor and control other aspects of the enrichment facility. The thermal guide 10 is formed of a relatively lightweight material so that it can be easily and quickly fitted to (and removed from) the UF6 cylinders 2. An example material is a cross-linked polymer, such as cross-linked polyethylene (e.g. PEX, PEX-Al-PEX and PERT), although alternative materials such as polybutylene could be used. The main body of the guide 10 may be bordered by a further heat insulating material at the heat insulating surface 14, such as a flexible microporous ceramics panel, in order to improve the thermally insulating properties of the heat insulating surface 14. An example method of using the thermal energy transfer apparatus 1 is described below with respect to FIG. 10. In a first step S1, a 48Y UF6 cylinder 2 containing UF6 which is of a natural or depleted concentration of U235 is received in a uranium material feed station of a uranium enrichment facility. The UF6 inside the cylinder 2 is in a solid state because the cylinder 2 has been stored at normal atmospheric temperatures of below 35° C. The UF6 is to be fed into an enrichment apparatus in which the UF6 must be in a gaseous state. In a second step S2, the 48Y UF6 cylinder 2 is wrapped in the thermal guide 10 of the thermal energy transfer apparatus 1. In the case of the 48Y cylinder 2, the thermal guide 10 comprises a plurality of sections 10A-F as described previously. The dimensions of the thermal guide 10 are matched to the length, diameter and circumference of the exterior of the 48Y cylinder 2 so that the thermal guide 10 fits around the cylinder 2 to surround it. The heat transfer surface 12 of the thermal guide 10 is in continuous contact with the exterior surfaces 7, 8 of the cylinder 2 to form a continuous thermally conductive contact patch around the cylinder 2 and over its ends. In a third step S3, the thermal guide 10 is connected to the heat transfer fluid circulation line 19. This allows heat transfer fluid 11 to flow from the circulation line 19 into the heat transfer medium containing region 13 of the thermal guide 10. The guide 10 may, for example, comprise a plurality of openings which are connectable to the circulation line 19 to receive heat transfer fluid 11 from the heat exchanger 18. In a fourth step S4, the temperature of the 48Y UF6 cylinder 2 is detected by the controller 20 using the temperature sensors 21 described previously. This allows the controller 20 to establish the amount of heating that will be required to convert the solid UF6 inside the 48Y cylinder 2 into a gaseous form. In a fifth step S5, the heat transfer fluid 11 is circulated around the looped heat exchange path 17 comprising the heat exchanger 18 and the thermal guide 10. This causes the heat transfer fluid 11 to pass from the heat exchanger 18 into the thermal guide 10 and back to the heat exchanger 18. In the thermal guide 10, the heat transfer fluid 11 is exposed to the temperature of the 48Y UF6 cylinder 2 through the thermally conductive heat transfer surface 12 of the guide 10. This causes heat exchange to take place between the heat transfer fluid 11 and the 48Y UF6 cylinder 2. Specifically, heat energy in the heat transfer fluid 11 conducts through the heat transfer surface 12 into the 48Y UF6 cylinder 2 and causes the temperature of the UF6 inside the cylinder 2 to increase. In a sixth step S6, the controller 20 continuously monitors the temperature of the 48Y UF6 cylinder 2 as the heat transfer fluid 11 is circulated. The controller 20 adjusts the level to which the heat transfer fluid 11 is heated in the heat exchanger 18 in order to obtain a target temperature for the cylinder 2 based on feedback from the temperature sensors 21. The controller 20 causes the heat exchanger 18 to heat the heat transfer fluid 11 to a temperature which is sufficient to continually raise the temperature of the 48Y UF6 cylinder 2. The rate at which the UF6 is heated may be varied by the controller 20, for example so as to cause an initial rapid rate of heating followed by a more gradual rate of heating as the UF6 cylinder 2 approaches the target temperature. In a seventh step S7, the controller 20 detects that the UF6 cylinder 2 has been heated to the target temperature. The target temperature is below the triple point of UF6 (64° C.), as previously described, but is sufficient for all of the UF6 inside the cylinder 2 to be in a gaseous state. In an eighth step S8, the thermal guide 10 is decoupled from the heat transfer fluid circulation line 19 and unwrapped from the 48Y UF6 cylinder 2. This involves releasing the quick release connectors 16, referred to previously, and may also involve draining the thermal guide 10 of heat transfer medium 11 so that it is lighter and easier to manipulate during removal from the UF6 cylinder 2. The thermal energy transfer apparatus 1 is now ready to be used to heat another 48Y cylinder 2 of UF6. Another example method of using the thermal energy transfer apparatus 1 is described below with respect to FIG. 11. In a first step M1, a 30B UF6 cylinder 2 ready to receive UF6 which has been enriched in its concentration of U235 is received in a uranium material take-off station of a uranium enrichment facility. The UF6 is fed into the cylinder 2 in a gaseous state because the UF6 has been enriched in a gaseous state in the enrichment apparatus. It is desirable to cool the UF6 in order to return it to a solid state. In a second step M2, the 30B UF6 cylinder 2 is wrapped in the thermal guide 10 of the thermal energy transfer apparatus 1. In the case of the 30B cylinder 2, the thermal guide 10 may comprises a single longitudinal section and two separate end sections, since the 30B cylinder 2 does not comprise the ribs 9 illustrated in the figures. The dimensions of the thermal guide 10 are matched to the length, diameter and circumference of the exterior of the 30B cylinder 2 so that the thermal guide 10 fits around the cylinder 2 to surround it. The heat transfer surface 12 of the thermal guide 10 is in continuous contact with the exterior surfaces 7, 8 of the cylinder 2 to form a continuous thermally conductive contact patch around the cylinder 2 and over its ends. The third step M3 is the same as that described above in relation to the first method. The thermal guide 10 is connected to the heat transfer fluid circulation line 19, which allows heat transfer fluid 11 to flow from the circulation line 19 into the heat transfer medium containing region 13 of the thermal guide 10. It will be appreciated that the second and third steps M2, M3 may be carried out before UF6 is fed into the cylinder 2 from the enrichment apparatus. In a fourth step M4, the temperature of the 30B UF6 cylinder 2 is detected by the controller 20 using the temperature sensors 21 described previously. This allows the controller 20 to establish the amount of cooling that will be required to convert the gaseous UF6 inside the 30B cylinder 2 into a solid state. The fifth step M5 is the same as the fifth step S5 described previously, apart from that the temperature of the heat transfer fluid 11 is lower, rather than higher, than the temperature of the UF6 cylinder 2. This causes heat energy in the 30B UF6 cylinder 2 to conduct through the heat transfer surface 12 into the heat transfer fluid 11 and causes the temperature of the UF6 inside the cylinder 2 to decrease. The sixth step M6 is also similar to the sixth step S6 described above. The controller 20 continuously monitors the temperature of the 30B UF6 cylinder 2 as the heat transfer fluid 11 is circulated, and the controller 20 may adjust the level to which the heat transfer fluid 11 is cooled in the heat exchanger 18 in order to obtain a target temperature for the UF6 cylinder 2 based on feedback from the temperature sensors 21. The controller 20 causes the heat exchanger 18 to cool the heat transfer fluid 11 to a temperature which is sufficient to continually lower the temperature of 30B UF6 cylinder 2. In a seventh step M7, the controller 20 detects that the 30B UF6 cylinder 2 has been cooled to the target temperature. The target temperature is sufficient for all of the UF6 inside the cylinder 2 to be in a solid state. In an eighth step M8, the thermal guide 10 is decoupled from the heat transfer fluid circulation line 19 and unwrapped from the 30B UF6 cylinder 2. The thermal energy transfer apparatus 1 is now ready to be used to cool another 30B cylinder 2 of UF6. The cylinder 2 shown in the figures is a 48Y UF6 cylinder 2, but it will be appreciated that, with the exception of the ribs 9, the features described with respect to the figures also apply to 30B UF6 cylinders 2 and other types of industry-standardized UF6 containers 2. Similarly, although the example methods and apparatus 1 have been described in the context of heating UF6 in a 48Y cylinder 2 and cooling UF6 in a 30B cylinder 2, the method steps and apparatus 1 could alternatively be used to heat or cool uranium material such as UF6 in any suitable uranium material container 2. For example, the method steps and apparatus 1 described above could be used to heat UF6 in a 30B cylinder 2 and/or to cool UF6 in a 48Y cylinder 2. The apparatus 1 has generally been described in the context of heating UF6 for supply to an enrichment apparatus and for cooling UF6 received from an enrichment apparatus. However, the method steps and apparatus 1 described above could alternatively, or additionally, be used to heat and/or cool UF6 in the cylinders 2 during UF6 blending operations to achieve a desired U235 concentration. The method steps and apparatus 1 could also be used to heat and/or cool UF6 cylinders 2 during UF6 recovery operations, for example in which UF6 is recovered from a damaged or outdated cylinder 2 and transferred into a new cylinder 2. The thermal energy transfer apparatus 1 described herein provides a heating and cooling process which is energy efficient. It also provides a process in which the temperature of the uranium material can be controlled accurately and in which desired changes to the temperature can be effected in a short period of time.
050341855	abstract	A control blade for a nuclear reactor having inserted upper end structural members and inserted lower end structural members connected to a plurality of wings each in the form of a generally rectangular plate having an longitudinal axis extending in the longitudinal direction of the control blade, the wing being disposed to form a cross-shaped section of the control blade. The wings and the structural members are connected to and supported on a central connection member. Each wing or a sheath member formed within each wing is formed from a diluted alloy obtained by diluting a long-lived neutron absorber such as hafnium with a diluent such as zirconium or titanium. A plurality of neutron absorber housing holes are formed in the diluted alloy section. Specifically, if the housing holes are formed to extend in the widthwise direction of the wing, the sectional area of each housing hole formed in a portion of the control blade corresponding to the region where the subcriticality in the reactor core becomes smaller after the reactor has been shut down by fully inserting the control blade into the core is larger than that of housing holes formed in the other regions. A structure having a means to cope with swelling of the neutron absorber can be applied to the former region.
059404647	summary	The present invention relates to tubes of zirconium-base alloy suitable for use, in particular, for constituting all or the outer portion of the cladding of a nuclear fuel rod, and also to a method of manufacturing them. Until now, use has been made above all of cladding made of a so-called "Zircaloy 4" alloy which contains tin, iron, and chromium in addition to zirconium. Numerous other compositions have been proposed, with content ranges that are often so broad that, to the person skilled in the art, they can be seen immediately to be purely speculative. In particular, various alloys have been proposed with a niobium content lying in a range so broad that their resistance to thermal creep is quite poor at maximum values, whatever the metallurgical treatments used in making the alloy. Alloys have also been proposed that contain, in addition to zirconium, tin to improve creep resistance, and iron. An object of the invention is to provide tubes that have simultaneously good creep behavior and good resistance to corrosion, even in a high temperature medium containing lithium, while nevertheless being capable of being manufactured with a low reject rate, and being suitable for use in making cladding or guide tubes for fuel assemblies. One of the causes of rejects is the formation of cracks during mechanical and heat treatments, giving rise to defects that make the tubes unacceptable. This risk exists particularly for high tin contents. To achieve the above objects, there is provided a tube of zirconium-base alloy containing, by weight, 0.8% to 1.8% niobium, 0.2% to 0.6% tin, and 0.02% to 0.4% iron, the alloy being in the recrystallized state or in relaxed state, depending on whether it is desired to enhance resistance to corrosion or to creep. The alloy has a carbon content lying in the range 30 parts per million (ppm) to 180 ppm, a silicon content lying in the range 10 ppm to 120 ppm, and an oxygen content lying in the range 600 ppm to 1600 ppm. The relatively high niobium content, which is always above the solubility limit (about 0.6%), provides high resistance to corrosion in an aqueous medium at high temperature. If used alone, niobium at such concentrations imparts creep characteristics to the alloy which are of interest but insufficient. Tin, when associated with niobium, improves creep resistance and also resistance to an aqueous medium containing lithium, without running the risk of causing cracks to be formed during rolling if its content does not exceed 0.6%. An iron content of up to 0.4% participates in compensating for the unfavorable effect of tin on generalized corrosion. The contents given above take account of the way in which tolerances and variations within a single ingot mean that the limits can be reached even for set specific contents lying within a narrower range. For example, set contents of 0.84% and 1.71% Nb may give rise within a single ingot to local contents of 0.8% and of 1.8% depending on proximity to the leading end or the trailing end of the ingot. In addition to the above-specified elements, the alloy contains inevitable impurities, but always at very low contents. It has been found that set content values of niobium in the range 0.9% to 1.1%, of tin in the range 0.25% to 0.35%, and of iron in the range 0.2% to 0.3% give results that are particularly favorable. Because of the relatively low tin content, recrystallization during metal-making can be performed at a relatively low temperature, below 620.degree. C., and that has a favorable effect on hot corrosion resistance and on creep. The invention also provides a method of manufacturing a tube for constituting cladding for a nuclear fuel rod or a guide tube for a nuclear fuel assembly. The initial alloy-making stage can be that performed conventionally for so-called "Zircaloy 4" alloys. However, the final stages are different, and in particular they make use of recrystallization heat treatments at relatively low temperature only. In particular, the method may comprise the following steps: making a bar of zirconium-base alloy having the above-specified composition; quenching the bar in water, after being heated to a temperature in the range 1000.degree. C. to 1200.degree. C.; drawing the bar into a tubular blank after heating to a temperature lying in the range 600.degree. C. to 800.degree. C.; annealing the drawn blank at a temperature in the range 590.degree. C. to 650.degree. C.; and cold-rolling said blank in at least four passes in order to obtain a tube, with intermediate heat treatments at temperatures in the range 560.degree. C. to 620.degree. C. The recrystallization ratio is advantageously increased from one step to the next in order to render grain size finer. In general, the final heat treatment is performed in the range 560.degree. C. to 620.degree. C. when the alloy is to be in recrystallized state, and in the range 470.degree. C. to 500.degree. C. when the tube is to be used in relaxed state. The alloy obtained in this way has resistance to generalized corrosion in an aqueous medium at high temperature, representative of conditions within a pressurized water reactor, that is comparable to that of known Zr--Nb alloys having high niobium content, and it has thermal creep resistance that is much greater than that of such alloys and that is comparable to that of the best "Zircaloy 4" alloys. By way of example, an alloy comprising 0.9% to 1.1% niobium, 0.25% to 0.35% tin, and 0.03% to 0.06% iron has been made. The metallurgical treatment sequence used comprised rolling over four cycles, with two-hour periods of heat treatment at 580.degree. C. interposed between the rolling step. The work hardening ratios and the recrystallization ratios were as follows: ______________________________________ Work hardening Recrystallization ratio (%) ratio (%) ______________________________________ First pass 40 70 Passes (2 or 3) 50 to 60 80 Last pass 30 100 ______________________________________ Additional tests have been carried out for determining the influence of the iron and tin content on alloys having 1% of niobium, with contents C, S.sub.i and O2 in the above indicated ranges formed into sheets and processed up to .SIGMA.a=5.23.times.10.sup.-18, with a final recristallization step at 580.degree. C. The corrosion tests were carried out: at 500.degree. C., 415.degree. C. and 400.degree. C. in water steam at 360.degree. C., in water containing 70 ppm of lithium.
052689395	abstract	A nuclear reactor is joined to a steam turbine by a main steamline for discharging steam thereto. A plurality of flow control valves regulate flow to the turbine, and a bypass valve selectively bypasses a portion of the steam around the turbine to its condenser. A pressure regulator and turbine controller are operatively joined to the control valves and the bypass valve for controlling steamflow to the turbine. An apparatus for detecting failure of one of the control valves is operatively connected to the bypass valve, and upon failure of one of the control valves to channel sufficient flowrate, the bypass valve is automatically opened to reduce reactor pressure rise. The failure detecting apparatus also provides a reduction demand signal for reducing reactor power for allowing the bypass valve to close.
	abstract	A fusion device produces fusion of neutral atoms and ions in an “aneutronic fusion” manner without neutrons as products utilizes strong ion-neutral coupling at high neutral densities. Ions and neutrals rotate together in a cylindrical chamber due to frequent collisions. High magnetic forces make the attainment of high rotation energy possible; the magnetic field in a medium can be set at very high values because of the absence of magnetic charges. The repeated acceleration by strong magnetic forces in the azimuthal direction makes possible very high ion velocity. Fusion takes place mainly between neutral particles. This approach can be applied to fusion with neutrons as well. Conventional fusion schemes and neutron sources can be realized using the principles described above in the generation of neutrals of high energies and densities.
	summary
	claims	1. A radiation phase change detection method for detecting a phase change of a radiation, the radiation phase change detection method comprising changing an interference state of the radiation using a phase grating configured to cause interference in the radiation radiated by a radiation source, a scintillator configured to convert the radiation into light, and a two-dimensional optical image pickup element,wherein the two-dimensional optical image pickup element is incapable of sampling a period of a self-image of the radiation generated through the phase grating, and is capable of sampling interference fringes generated between the period of the self-image and a period of a pixel pitch of the two-dimensional optical image pickup element. 2. A radiation phase change detection method according to claim 1, further comprising placing an object between the radiation source and the phase grating, or between the phase grating and the scintillator. 3. A radiation phase change detection method according to claim 1, further comprising arranging the two-dimensional optical image pickup element so that, when the period of the self-image is defined as D1, and the pixel pitch of the two-dimensional optical image pickup element is defined as D2=kD1, k falls in a range of ½<k≤3/2. 4. A radiation phase change detection method according to claim 3, further comprising arranging the two-dimensional optical image pickup element so that the interference fringes formed by D1 and D2 have a period of 2 times D2 or more and 100 times D2 or less. 5. A radiation phase change detection method for detecting a phase change of a radiation, the radiation phase change detection method comprising:arranging a two-dimensional optical image pickup element, which includes a scintillator, so that, when a period of a self-image generated through a phase grating is defined as D1, and a pixel pitch of the two-dimensional optical image pickup element is defined as D2=kD1, k falls in a range of ½<k≤3/2, and so that interference fringes formed by D1 and D2 depending on a relationship in arrangement of the two-dimensional optical image pickup element with respect to the self-image have a period of 2 times D2 or more and 100 times D2 or less;acquiring a phase change of the interference fringes before and after insertion of an object; andoutputting an image on a phase change of the radiation caused by at least the insertion of the object. 6. A radiation phase change detection method according to claim 5, further comprising adjusting the relationship in arrangement of the two-dimensional optical image pickup element with respect to the self-image through rotation so that the interference fringes have a period of 100 times D2 or less. 7. A radiation phase change detection method according to claim 5, wherein the scintillator comprises a scintillator capable of resolving 100 lp/mm at least with a thickness of 150 μm. 8. A radiation phase change detection method according to claim 5, wherein the scintillator comprises a eutectic phase-separated scintillator, in which a plurality of fiber structures containing GdAlO3 are surrounded by a material containing Al2O3. 9. A radiation phase change detection method according to claim 5, further comprising arranging a fiber optic plate, which has a periodicity of half the period D1 of the self-image or less, between the scintillator and the two-dimensional optical image pickup element. 10. A radiation imaging apparatus, comprising:a radiation source;a phase grating configured to cause interference in a radiation radiated by the radiation source;a scintillator configured to convert the radiation into light; anda two-dimensional optical image pickup element,wherein the two-dimensional optical image pickup element is incapable of sampling a period of a self-image of the radiation generated through the phase grating, and is capable of sampling interference fringes generated between the period of the self-image and a period of a pixel pitch of the two-dimensional optical image pickup element. 11. A radiation imaging apparatus according to claim 10, wherein the two-dimensional optical image pickup element is arranged so that, when the period of the self-image is defined as D1, and the pixel pitch of the two-dimensional optical image pickup element is defined as D2=kD1, k falls in a range of ½<k≤3/2. 12. A radiation imaging apparatus according to claim 10, wherein the two-dimensional optical image pickup element is arranged so that the interference fringes formed by D1 and D2 have a period of 2 times D2 or more and 100 times D2 or less. 13. A radiation imaging apparatus, comprising:a radiation source;a phase grating;a scintillator; anda two-dimensional optical image pickup element,wherein the radiation imaging apparatus is configured to detect a phase change of a radiation by:arranging the two-dimensional optical image pickup element so that, when a period of a self-image generated through the phase grating is defined as D1, and a pixel pitch of the two-dimensional optical image pickup element, which includes the scintillator, is defined as D2=kD1, k falls in a range of ½<k≤3/2, and so that interference fringes formed depending on a relationship in arrangement of the two-dimensional optical image pickup element with respect to the self-image have a period of 2 times D2 or more and 100 times D2 or less;acquiring a phase change of the interference fringes before and after insertion of an object; andoutputting an image on a phase change of the radiation caused by at least the object. 14. A radiation imaging apparatus according to claim 13, wherein the radiation source comprises an X-ray source. 15. A radiation imaging apparatus according to claim 13, further comprising at least one selected from the group consisting of a driving system, an arithmetic section, and an image acquisition unit.
055457975	description	EXAMPLE 1 Preparation of Zr.sub.1-x Pu.sub.x SiO.sub.4 in the Laboratory Dissolve Pu in hydrochloric acid (HCl) and add nitric acid (HNO.sub.3. Evaporate off HCl at about 100.degree. C. Add more HNO.sub.3 and evaporate again (if necessary, repeat to dissolve Pu completely). Add HNO.sub.3 and dilute with water to form an aqueous Pu-nitrate solution. Mix stoichiometric quantities of the Pu-nitrate solution, zirconium nitrate [Zr(NO.sub.3).sub.4 .multidot.yH.sub.2 O] and tetraethylorthosilicate (TEOS) in ethyl alcohol and water to achieve the desired loading of Pu (x-value). Add gadolinium nitrate [Gd(NO.sub.3).sub.2 .multidot.yH.sub.2 O] solution in a small quantity, if a neutron poison is necessary for criticality control. Gd will partially substitute for Pu or Zr. At this stage, a .UPSILON.-radiation emitter, e.g., Co-60, can also be added in small quantities, if easy physical access to the final waste form is to be prevented. Heat this solution to 40.degree. to 50.degree. C. for several days to allow nucleation to occur. Add ammonium hydroxide (NH.sub.4 OH) to form a precipitate. Remove the precipitate and dry at about 90.degree. to 100.degree. C. Calcine the dried precipitate at about 800.degree. C. to remove residual water and to decompose nitrate. The powder can be processed into a final waste form by cold pressing and subsequent sintering at about 1800.degree. C. to produce a high density, impervious, and chemically durable solid that can be placed in a metal canister for transportation, storage, and final disposal in a geologic repository. Alternatively, a metal-sheathed high-density waste form can be obtained by uniaxially cold pre-pressing and subsequently hot pressing the powder in a metal bellows container at temperatures from 1150.degree. to 1350.degree. C. These bellows can then be placed in metal containers for transportation, storage, and disposal. EXAMPLE 2 Preparation of Zr.sub.1-x Pu.sub.x SiO.sub.4 at a Larger Scale For obvious reasons, this process has not yet been tested but is envisaged to be conducted as follows: Convert Pu metal to Pu-nitrate as in Example 1 and dry the nitrate at 90.degree. to 100.degree. C., or convert Pu metal to Pu-oxide by oxidation in air or in oxygen. The rate of oxidation can be controlled by the amount of oxygen or air in the reaction cell. Mix stoichiometric quantities of Pu-oxide with Zr-oxide (ZrO.sub.2) and silicon oxide (SiO.sub.2) powders to achieve desired waste loading. Add neutron poison as an oxide (Gd.sub.2 O.sub.3) powder, if desirable. If Pu is added as nitrate, calcine the mixture at 650.degree. C. to remove water and decompose nitrate. Intimate mixing of the powders, e.g. in a screw blender, is necessary to facilitate the solid state reaction and to keep the reaction temperatures and pressures as low as possible. If necessary, ZrO.sub.2 and SiO.sub.2 powders could be obtained by hydrolysis of mixtures of respective organic precursors (e.g., TEOS or TMOS for SiO.sub.2). Amorphous silica or other reactive products such as xerogels can also be used. After mixing, the powder can be processed as described above. The powder must be transferred into a bellows feeder from where it can be vibrated into the bellows. Prior to cold pressing a small quantity of zircon (ZrSiO.sub.4) doped with a .UPSILON.-emitter, or the .UPSILON.-emitter as such, could be added, if desirable. The .UPSILON.-doped zircon need not be distributed homogeneously and can be introduced into the feeder or into the bellows. Hence, only the steps of bellows filling and cold and hot pressing are conducted in a shielded environment. The exact conditions of the hot pressing step (temperature, pressure and time) of large scale processing of the waste form, starting with oxide mixtures, depend on the details of the process. Approximate values should be as follows: Temperatures 1150.degree. to 1350.degree. C., pressure 15-30 MPa, time 1 to 2 hours. It should be noted that the cold pressing and the heating are carried out in the same bellows. The bellows are first cold pressed to increase the thermal conductivity of the powder, whereupon heating is effected. This will decrease the reaction time at temperature and under pressure. From the foregoing, it can be seen that the inventive method offers a number of advantages over the heretofore known methods. For example, due to the relatively high waste loading that is possible as well as the smaller volume that is achieved, deep, permanent disposal of plutonium in geologic environments where a borosilicate waste form glass would not be stable is possible. Furthermore, due to the high durability of the zircon structure, disposal in an open system in which ground water is present is also possible. The reason that zircon is an improvement over glass for deep disposal is threefold. First of all, zircon is stable at higher temperatures, and deep disposal brings glass into a temperature range in which it is not stable due to the geothermal gradient. Secondly, a higher waste loading in zircon is possible, and this reduces the volume of material that must be placed down a drill hole. Higher waste loading is possible pursuant to the present invention because zircon is durable at high temperatures and the low release rate due to chemical corrosion means that the probability of release, concentration, and ultimately criticality is minimized. Thirdly, methods of criticality control of PuO.sub.2 are well-known from mixed-oxide full (MOX) fabrication and can be applied to the fabrication of zircon. In summary, due to the extremely durable phase of the zircon structure the latter can be used as a plutonium host for the disposal of large quantities of plutonium. This long-term durability of the zircon structure has been confirmed from natural occurrences in diverse and extreme geologic environments over extremely long periods of time. The very low solubility of the zircon structure ensures that plutonium will not be concentrated by later cycles of geochemical alteration to values that might lead to criticality. Finally, the lower volume provided by the inventively produced waste product, and the greater durability, particularly at elevated temperatures, expands the range of possible geologic disposal sites. The present invention is, of course, in no way restricted to the specific disclosure of the specification and examples, but also encompasses any modifications within the scope of the appended claims.
043022944	description	The fuel assembly shown in FIG. 1 comprises a set of parallel fuel rods 1 held in a rigid bundle by means of transverse cross-pieces 2, which are arranged at approximately regular intervals over the length of the rods, and end plates 3. As shown in FIG. 2 each cross-piece 2 defines cells 4 through which the fuel rods extend. Some of these cells 4 are occupied by support tubes 5 which are substituted for a certain number of fuel rods. The length of the support tubes 5 is slightly greater than the length of the fuel rods 1, and the tubes 5 are joined at their ends to grids 6. As can be seen in FIG. 3, each grid 6 provides a network of square cells 10. When the grid 6 is in position in the assembly, as shown in FIG. 1, the support tubes 5, which are permanently fixed to the grids 6 either by crimping or by crimping and welding, are positioned in some of the cells 10. Each of the grids 6 comprises an assembly of metal strips or small metal plates 11, of a certain height, which intersect and are fixed to one another. Some of the cells of each grid 6, which cells 10 are indicated by crosses 12 in FIG. 3, are intended to receive sockets for joining the grid to the respective end plate 3. The network consisting of the cells 10 of each grid 6 is arranged, relative to the fuel elements of the assembly, in such a way that each of the cells 12 inside which support tubes are not arranged is in line with a fuel element 1. The network of the grid 6 is thus arranged in substantially the same manner as the cross-piece shown in FIG. 2. Furthermore, the size of the cells 10 of each of the grids 6 is such that the fuel elements 1 can be extracted by displacing them longitudinally and passing them through the cells of the grid. When one of the end plates 3 is dismantled, it is therefore possible to extract the fuel elements from the assembly, even though the grids 6 remain permanently fixed to the support tubes 5. Reference will now be made to FIGS. 4, 5, 6 and 7 in order to describe the method of fixing the grids 6 to the end plates 3. The method is described in respect of the upper grid and end plate. The fixing of the lower grid to the lower end plate is effected in a corresponding way. This fixing is achieved by means of sockets 7, of cylindrical shape, each of which possesses, at its upper end, an enlarged part 14 of which the lower face constitutes a shoulder 15 which can come into contact with the bottom of a cavity 16 of circular section which is provided in the outer upper face of the end plate 3. When the upper end plate 3 is positioned on the assembly, its lower face comes into contact with the upper face of the grid 6, and passages 17 of square section, which are provided in the end plate 3 at the centers of the recesses 16 and pass right through this plate 3, coincide with the square cells 10 of the grid. The dimensions of the transverse sections of the recesses 17 are approximately the same as the dimensions of the cells 10. When the socket 7 is in position it extends through a passage 17 and a cell 10 of the grid, as shown in FIG. 5 and its shoulder 15 comes to rest on the bottom of the recess 16 provided on the upper face of the end plate. The diameter of the cylindrical central part of circular section of the socket 7 is slightly smaller than the side of the cell of the grid. The socket 7 is also provided, at its lower part, with an enlarged part 18 of square transverse section, the dimensions of which are slightly smaller than the dimensions of a cell 10 and of a passage 17, so that the part 18 can, when appropriately orientated, pass through a passage 17 and a cell 10. The enlarged upper part 14 of the socket 7 is provided with a slot 20 for engagement by the blade of a screwdriver in order to orientate the socket by rotating the screwdriver. It is thus possible to arrange the socket 7 in such a way that the enlarged lower part 18 is in the position shown in FIG. 7, in which the corners of the part 18 come to bear on the lower surface of the grid. Once the socket 7 has been introduced into the end plate 3 and the grid 6, as shown in FIGS. 5 and 7, the socket is fixed, with pressure, against the end plate and the inner surface of the grid by expanding the cylindrical part of the socket 7, at the level of the square section passage 17 in the end plate, into the corners of the passage 17, as shown by 19 in FIG. 6. The deformation 19 can be obtained inside the socket 7 by expansion, that is to say by means of an elongate tool comprising, at its end, rollers which deform the metal of the socket. During this deformation, the metal which is pushed out penetrates into the four corners of the square passage 17 and thus prevents the rotation of the socket relative to the end plate and to the grid. The deformation also causes a slight shortening of the socket, and this causes the end plate 3 to be tightened against the grid 6 via the parts 15 and 18 of the socket, which are in contact with the upper part of the end plate and the lower part of the grid respectively. It will be understood that the unit consisting of the support tubes 5, the cross-pieces 2 and the grids 6 which are fixed to the ends of the support tubes retains a certain rigidity when one or both of the two end plates 3 are removed, and that it is then possible to extract fuel elements from this structure, or introduce them into this structure, in order to replace the elements or to remove them for the purpose of checking or testing operations. When it is desired to fit the assembly together again, the fixing of the or each end plate to the corresponding grid is extremely easy because it suffices to place the end plate on the grid so that the passages 17 are in alignment with the cells of the grid, and then to introduce the fixing sockets 7 into some of the passages and cells and to orientate the sockets by means of a screwdriver engaging the slots 20, and finally to produce the deformations 19 inside the sockets 7. Conversely, when it is desired to dismantle the assembly in order to remove a defective element or an element which is intended for checking or testing operations, the assembly then being located in the swimming pool of the reactor, the operations are even simpler and can easily be carried out by remote control with the assembly immersed. In order to dismantle the assembly, it suffices, in fact, to introduce the blade of a screwdriver successively into the slots 20 of the various sockets 7 arranged in the end plate which it is desired to dismantle, and to rotate each socket using the screwdriver. The protuberances 19, made in the socket when it was mounted, are then deformed by contact with the inner surface of the passage 17, and this then permits the rotation of the socket and the release of the tightening force between the end plate and the grid. The sockets can then easily be extracted from the cavities, and the end plate can be removed. The fuel elements are then accessible again. It is seen that the main advantage of the above described assembly is that it has a connection between the end plates and the rest of the assembly, the making and breaking of which is extremely simple, and, in particular, which avoids machining operations, such as grinding or cutting, when the assembly is dismantled, which operations are likely to contaminate parts of the fuel assembly, likewise, any welding operation which presents the same contamination risks. Furthermore, even when the end plates have been dismantled, the assembly retains a certain rigidity and a certain stability. The invention is not intended to be restricted to the embodiment which has been described but, on the contrary, includes all variants thereof, and points of detail can be modified without thereby going outside the scope of the invention. For example, the upper part of the socket 7 may be provided with a hexagonal or square recess into which it is possible to insert a key, of corresponding shape, in place of a slot for a screwdriver blade. The socket may be tubular, as shown in FIGS. 5 and 6, or it may have solid parts obstructing the flow through the end plate. The deformable part of the socket may be located at any point on the socket or it may consist of a part added to the socket. The lower part of the socket may have one of a number of very diverse shapes, provided that the lower part can penetrate through the passage in the end plate and the cell of the grid, and can then, after rotation, present bearing surfaces to the inner surface of the grid. It is also possible to join one or both of the grids to the upper cross-piece or to the lower cross-piece respectively, that is to say to the cross-piece located closest to the upper end or to the lower end of the assembly, depending on whether the upper grid or the lower grid is concerned. The grid can even form part of the upper cross-piece or the lower cross-piece if the height of the cross-piece is increased and if openings are provided therein for the passage of the locking ends of the sockets. The grid may be joined to the cross-piece by a number of strips and small plates of which it is made, and particularly by the small plates arranged at the exterior of the grid. The sockets 7 and the grids 6 are generally made of stainless steel of the same type as the other parts of the assembly, but the use of other materials, chosen, for example, for their useful mechanical properties, is not excluded. However, the grids are preferably made of a material which is identical to the material constituting the end plates, so as to avoid problems due to the thermal expansion of these pieces. Finally, the connection, by means of a grid and sockets, between the assembly and the end plates can be used for the upper end plate or for the lower end plate of the assembly, or for both these end plates.
043538630	claims	1. Method for localizing a leaking rod inside a nuclear fuel assembly, in which for each rod in an assembly, the radio-activity in at least two discrete rod rows in which said rod lies is measured and a leaking rod is localized by sensing a lowering of the radio-activity in the tested rows where said rod lies relative to the radio-activity in an identical row of non-leaking rods. 2. Method as defined in claim 1, in which the radio-activity of the fission products accumulated in the rod plenum is measured. 3. Method as defined in claim 1, which comprises measuring the radio-activity from a tracer added to said rods during the manufacturing thereof. 4. Method as defined in claim 1, which comprises making use of a change in the ratio between two or more radio-active products to localize said leaking rods. 5. Method as defined in claim 1, in which the .gamma.-radiation generated by said fission products is measured. 6. Method as defined in claim 1, which comprises measuring the radio-activity at the level of the rod plenums. 7. Method as defined in claim 1, which further comprises measuring the radio-activity of assembly rods inside a desactivating or storage tank which contains a cooling medium, or inside a cased cell. 8. Method as defined in claim 1, in which the cooling medium is discharged from the area of those rods to be subjected to radio-activity measuring. 9. Method as defined in claim 1, in which use is made of releasable assemblies to allow replacing said leaking rods. 10. Method as defined in claim 1, which further comprises subjecting a radiation sensor and a rod assembly to a relative displacement along a cross-wise direction, preferably substantially at right angle to the rod axis, to allow measuring with said sensor, the radio-activity from one rod row at a time. 11. Method as defined in claim 10, in which said displacement is comprised of a continuous translating of the assembly and/or sensor. 12. Method as defined in claim 10, in which said displacement is comprised of a stepwise translating over a distance which is substantially equal to the spacing between the axial planes of two succeeding rod rows in said assembly along a cross-wise direction, preferably at right angle to the rod axis. 13. Method as defined in claim 10, in which said displacement is comprised of a co-ordinated rotating of said rod assembly and said sensor. 14. Method as defined in claim 1, in which one sensor is used for every rod row, said sensors being aligned with the rods along one or more side surfaces from said assembly. 15. Method as defined in claim 1, which further comprises heating or letting the rods get heated before said radio-activity measuring. 16. Method as defined in claim 1, which comprises generating an underpressure around said rods. 17. Equipment for localizing a leaking rod inside a nuclear fuel assembly, notably for the working of the method as defined in claim 1, which comprises a tank enclosing a cooling medium into which dips partly at least a rod assembly, and at least one radio-active radiation sensor arranged behind a collimator with such a size that only the rod plenums lie in the sensing or viewing area of said sensor(s). 18. Equipment as defined in claim 17, in which the sensor and the assembly are so mounted as to be movable relative to one another along a cross-wise direction, preferably at right angle to the rod axis. 19. Equipment as defined in claim 17, in which said collimator is provided in the side wall of the tank or cell, facing the plenums from the assembly rods. 20. Equipment as defined in claim 17, in which the unit formed by the sensor and the collimator is immersed inside the tank. 21. Equipment as defined in claim 17, in which said collimator is provided facing the sensor, with a slit the width of which is smaller than the outer rod diameter, for example about 80% thereof. 22. Equipment as defined in claim 17, which further comprises a cylinder to be arranged above the rod assembly and dipped partly at least into the tank cooling medium, said cylinder being connected through the top thereof, to a pressurized gas supply to allow forcing the cooling medium away. 23. Equipment as defined in claim 22, in which said cylinder is open at the bottom thereof, can be arranged above the rod assembly and can dip partly at least into the tank cooling medium. 24. Equipment as defined in claim 22, in which said cylinder is connected through the top thereof to a pressurized gas supply or a vacuum pump, and it is provided at the bottom thereof with closing means.
053435060	summary	CROSS-REFERENCE TO RELATED APPLICATION This application is a Continuation of International Application Ser. No. PCT/DE91/00993 filed Dec. 18, 1991. BACKGROUND OF THE INVENTION Field of the Invention The invention relates to a nuclear reactor installation, in particular for light water reactors, with a reactor pressure vessel having a reactor core and a core catcher device. Such a nuclear reactor installation is known from U.S. Pat. No. 3,607,630. In addition, that known nuclear reactor installation has the following features: a supporting and protective structure delimits a reactor cavern with a bottom region and a circumferential wall, and the reactor pressure vessel, disposed in the reactor cavern at vertical and lateral distances in relation to the bottom region and the circumferential wall, is seated in the supporting and protective structure. In that case the core catcher device has a collecting basin for the core melt, which can be cooled by means of a coolant and is embedded within the reactor cavern and below the reactor pressure vessel in the bottom region of the supporting and protective structure. The collecting basin, which is also known as a "core catcher", is flat, pan-shaped and water-cooled internally. It is connected through an ascending pipe to a flood container disposed at a higher level. The wet steam forming in a hypothetical case of a core melt, i.e. when the core melt is distributed in the collecting basin, is blown off through outlet lines into the containment vessel or into condensation devices (steam separators). The condensed cooling water is returned to the flood container. The collecting basin is formed of a plurality of parallel tubes connected at the respective inlet and outlet ends to a common distribution or collection tube. However, the relatively good cooling properties of such a known collecting basin can be impaired, particularly in the case of nuclear reactors with higher output, if the tube structure of the collecting basin is deformed by large falling masses and the cooling cross sections are reduced or blocked by thereby. The core catcher according to Published UK Application No. GB 2 236 210 A, has a collecting basin in "multilayer sandwich construction", with a bearing, downwardly arched steel pot, cladding of interlocked zirconium blocks and a steel skin covering the cladding at the top, which melts through in the case of an impacting core melt, i.e. it is sacrificed. Since the catch volume of the core catcher is relatively small and, alternatively to gas cooling, is only provided with a standing water column in the shield pit or reactor cavern, effective continuous cooling of the core melt (which in the beginning may have temperatures above 2000.degree. C.) would only be possible with small reactor outputs, since otherwise film boiling could occur at the outer steel jacket of the core catcher, along with the danger of considerably reduced heat transfer. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a nuclear reactor installation with a core catcher device and a method for exterior cooling of the latter by natural circulation, which overcome the hereinafore-mentioned disadvantages of the heretofore-known devices and methods of this general type and in which it is possible, by means of its collecting basin construction and support, to assure sufficient cooling channel cross sections and cooling of a possible core melt, even with increased reactor output and reactor core weights, without having to fear impairment of the structure defining the cooling channels by the deformation forces of impacting masses. In addition, within the framework of subordinate and coordinate objects, it is intended in connection with a nuclear reactor installation according to the invention to provide requirements for cooling the collecting basin in accordance with the principle of natural circulation with a liquid and, in addition, to provide a dual cooling system (air cooling and water cooling) in such a way that, in the case of emergency cooling, air cooling is replaced, at least partially, by water cooling without special switching commands. Another subordinate object is effectively keeping away the radioactive radiation emanating from the bottom of the collecting basin in case of a core melt, from the wall sections located above the collecting basin of the supporting and protective structure. A further subordinate object is integrating a heat insulation surrounding the reactor pressure vessel into the system being formed of the collecting basin and dual cooling device. Heretofore there has been no lack of suggestions for eliminating an incident of an accidental core melt through special safety measures. The recently developed safety philosophy assumes that it is better from the viewpoint of safety technology to include a core melt incident in the considerations, even though the possibility of its occurrence might be infinitely small. The invention is based on this viewpoint. A particularly effective protective barrier for preventing undesirable consequences of a core melt incident is intended to be provided by means of the invention. Additional subsidiary objects connected with the above defined general objective ensue from the following considerations. It is desirable in connection with light water nuclear reactors in general, and with pressurized water nuclear reactors in particular, that the integrity of the containment be maintained in all assumed incidents, i.e. also in case of a core melt, regardless of whether it is a beginning partial core melt or a complete melt-through of the core. The following requirements in particular are set forth for controlling such an incident: a. no fission products must be allowed to escape in large amounts from the core melt into the containment; instead, the core melt must remain covered by continuously cooled water (or another suitable liquid coolant), or a crust must be formed by means of cooling in order to achieve a catching effect; b. the core melt must not be allowed to come into interaction with the concrete of the support structure of the safety container, at least not during the initial days of the event which exceeds the design specifications. That is also true because otherwise hydrogen, water vapor, non-condensable gases and other reaction products could be released; c. cooling of the core melt over a long period of time must be assured, by means of which the post-decay heat is transferred to a heat sink and the melt is caused to harden and is maintained in a solid aggregate state over a long period of time; and d. large-size steam explosions, which can occur when large amounts of core melt mass fall or "plop" into a water bath, must be prevented. With the foregoing and other objects in view there is provided, in accordance with the invention, a nuclear reactor installation, in particular a light water reactor installation, comprising a reactor pressure vessel, a reactor core in the reactor pressure vessel; a supporting and protective structure supporting the reactor pressure vessel and surrounding the reactor pressure vessel on the bottom and laterally, the supporting and protective structure having a bottom region and a circumferential wall; a core catcher device for the reactor core having a collecting basin for a core melt, which can be cooled by means of a cooling liquid and which is installed below the reactor pressure vessel, the collecting basin having a bottom wall and a jacket wall being respectively separated from the bottom region and the circumferential wall of the supporting and protective structure by a spacing gap; cooling channels disposed in the spacing gap at the bottom and on the sides for exterior cooling of the collecting basin with a cooling liquid; and turbulence bodies disposed in a surface region of the bottom wall for generating a turbulent flow of the cooling liquid flowing from the inside to the outside over the bottom wall toward the jacket wall. Advantageous further embodiments of the subject of claim 1 are recited in dependent claims 2 to 21. The advantages which can be attained by means of the invention are to be found mainly in the following: the collecting basin has such a height (at least approximately 3 m) that the minimum height for forming a naturally circulating flow is provided in liquid-filled cooling channels on the bottom and the sides (outer cooling system). The collecting basin protects the concrete of the supporting and protective structure (biological shield) not only with its bottom wall, but also with its upwardly extending jacket wall, against the effects of heat and radiation emanating from the reactor pressure vessel or a core melt. In this case the inner width (inner diameter) and the depth of the reactor cavern is suitably made sufficiently large so that, even with a sufficiently great spacing gap (=gap width of the outer cooling system), the collecting basin encloses a volume which permits the lining of the interior of the base body of the collecting basin, preferably a crucible made of a temperature-resistant steel alloy, with a protective layer and with masonry of shielding concrete blocks, while still providing sufficient receiving space for the possible core melt case. The crucible-like base body and the support by the turbulence bodies on its underside, in which case the turbulence bodies are in the form of turbulence-generating flow guidance bodies, can be easily constructed sufficiently strong and with a load-bearing capacity distributed over the base so that sufficiently large cooling cross sections can be maintained, even under great dynamic and static loads. Based on the large flow-through cross sections of the exterior cooling system, the naturally circulating flow with a corresponding coolant flow rate which can be generated and the generated turbulent flow, it is possible to make the exterior cooling of the collecting basin so effective that film boiling at the exterior cooling surfaces of the collecting basin can be prevented even under the greatest thermal loads. In accordance with another feature of the invention, the cooling channels at the bottom are connected through an inlet channel configuration, and the cooling channels on the jacket are connected through an outlet channel configuration, to a cooling water reservoir provided outside of the supporting and protective structure and forming a reactor housing sump or being connected therewith with such lift that, with a hot collecting basin and water-filled cooling channels, a naturally circulating flow through the cooling channels is generated. The collecting basin can be seated while being suspended from the supporting and protective structure. For this purpose it can be provided, similar to a core container seated and suspended within a reactor pressure container, with a support flange, by means of which it is seated on corresponding support surfaces of the supporting and protective structure. However, the collecting basin is preferably seated on the bottom part of the supporting and protective structure by means of the turbulence bodies (which are then also support bodies), and a dual function (support and turbulence generating) can be achieved in this way. In order to allow for the unhindered heat expansion in the radial direction, the bottom wall of the collecting basin can be seated glidingly and/or resiliently on these support bodies, or the latter can be seated in this manner on the bottom region of the supporting and protective structure. In accordance with a further feature of the invention, the collecting basin is a crucible and for this purpose its bottom wall is curved towards the bottom or the outside, wherein its bottom wall merges into the jacket wall through a rounded-off edge area, and the jacket wall preferably tapers slightly conically from the rounded-off edge area to the upper edge of the collecting basin. In accordance with an added feature of the invention, the bottom wall of the collecting basin widens in the shape of a flat envelope of a cone from the lowest central area to the edge area, and the intersecting surfaces of which, located in axial-radial intersecting planes, extend with a slight angle of slope relative to the horizontal. This slight inclination and the rounding-off in the edge area make the bathing of the bottom and jacket walls of the collecting basin with cooling liquid, especially with water, easier in accordance with the natural circulating principle and in this way make effective cooling possible. In accordance with an additional feature of the invention, in order to provide dynamically balanced even cooling of the collecting basin, an inlet channel configuration discharges into the cooling channels in the central area of the bottom wall of the collecting basin through an inlet chamber, the cooling channels on the bottom extend outwardly from the inlet chamber as far as the edge area of the collecting basin, and an upwardly extending cooling channel adjoins the edge area on the jacket side and terminates in the outlet channel configuration. In accordance with yet another feature of the invention, the inlet channel configuration penetrates through the bottom region of the supporting and protective structure and extends from the bottom wall of a chamber forming the outer cooling water reservoir as far as the central area of the bottom wall of the collecting basin. Accordingly, the outlet channel configuration penetrates through the circumferential wall of the supporting and protective structure, forms a continuation of the cooling channel on the jacket and terminates in the area of the upper level of the cooling water reservoir. In accordance with yet a further feature of the invention, in connection with the protective barrier function of the collecting basin, the base body of the collecting basin is formed of a crucible being formed of a non-corroding, temperature-resistant steel alloy, the interior bottom and jacket surfaces of the crucible are lined with a protective shell used for protecting the crucible material against attacks by the melt, and a sacrificial material deposit follows the protective shell as a second protective layer on the crucible, the amount of which is sufficient for reacting with the maximally possible volume of core melt entering the collecting basin in case of a possible incident. In accordance with yet an added feature of the invention, the protective shell is formed of one of the following alloys, either singly or in combination: MgO, UO.sub.2 or ThO.sub.2. In accordance with yet an additional feature of the invention, the sacrificial material deposit is a masonry structure of shielding concrete blocks. Lining with a deposit of sacrificial material in the form of a granulate or an even more large-grained bulk material or preferably in the form of a masonry facing of shielding concrete blocks serves the purpose of altering the material values of the mixture in a directed way, for example for the purpose of: protecting the wall of the collecting basin against high temperatures immediately following the penetration of the core melt into the collecting basin; using up energy through melting the sacrificial material, with the result of delaying heating-up of the melt and thus being in a position of expecting lower values for the post-decay heat for cooling; making the core melt more fluid; increasing its heat conductivity; increasing its surface; improving the heat transfer from the core melt to the cooling surfaces; preventing steam explosions by displacing volumes of water; providing defined calculation bases by means of the known properties of the sacrificial material; and lowering the melting point of the mixture and the temperature of the melt. In accordance with again another feature of the invention, the channel bodies mentioned above with regard to their property as flow guidance bodies for generating a turbulent flow in the exterior cooling system, are shaped as so-called delta wings in the form of prisms with three-sided surfaces and are fastened at least on the bottom of the supporting and protective surface located opposite the bottom wall with the cooling gap of the collecting basin. Such delta wings have proven themselves to be particularly effective for generating a turbulent flow in the cooling gap. The delta wings assist in preventing a film of steam on the underside of the plate through which the heat transfer number, which is decisive for heat transfer from the heated plate to the cooling water flow, would be undesirably reduced. The natural circulation in the cooling gap can be intensified by means of the generated turbulent flow in such a way, that it is possible to maintain a sufficient safe distance from the so-called critical heating surface load. In accordance with again a further feature of the invention, the channel bodies are also used for support if they are constructed as pipe sockets, and the pipe sockets are provided on their ends facing the bottom wall portion of the collecting basin with channel recesses for generating partial cooling water flows, so that the latter also bathe the bottom wall in the area of the pipe sockets. These pipe sockets can be constructed either as simple flow guidance bodies or as turbulence-generating flow guidance bodies. In accordance with again an added feature of the invention, when the pipe sockets are constructed as turbulence-generating flow guidance bodies, two respective U-shaped channel recesses aligned in the flow direction are provided per pipe socket and the ends thereof are made angular to increase turbulence. As was already explained, the collecting basin also has a radiation shielding function. In accordance with again an additional feature of the invention, the radiation shielding system provided thereby is completed in an advantageous manner by installing a shielding ring above the collecting basin and adjoining it in the annular chamber between the circumferential wall of the supporting and protective structure and the outer periphery of the reactor pressure vessel. In particular, the shielding ring assumes the function of the biological shield in the circumferential area of the reactor core at those places where the circumferential wall (biological shield) is penetrated by outlet channels, so that the radioactive radiation emanating from the reactor core is kept away from the spaces outside the supporting and protective structure. Suitably the shielding ring is formed of shielding concrete, which is also called leca-concrete. The shielding ring has a wall thickness closely approximating the wall thickness of the biological shield (supporting and protective structure) and preferably its extent in height is somewhat greater than its wall thickness. It is also advantageous to incline the top of the shielding ring, so that a larger annular surface is provided as the outlet cross section for air cooling channels. In accordance with another feature of the invention, the shielding ring is anchored or braced on the circumferential wall of the supporting and protective structure. In accordance with a further feature of the invention, the shielding ring is formed of prestressed concrete in particular, and its steel reinforcement is preferably combined into a uniform steel reinforcement system together with the steel reinforcement of the supporting and protective structure which also is formed of prestressed concrete. The shielding ring can be poured on site, in which case an appropriate masonry work has to be provided, or it can be assembled from individual ring segments which are pre-fabricated. In the latter case the ring segments of the shielding ring are advantageously interlocked with each other and with the circumferential wall of the supporting and protective structure. In accordance with an added feature of the invention, the exterior cooling system of the collecting basin is constructed as a dual air and water cooling system which, during the normal operation of the nuclear reactor installation, i.e. when the exterior cooling system is dry, is used for air cooling of the nuclear reactor pressure vessel or of the outside of a thermal insulation enclosing it. For this purpose the inlet channel configuration is connected with a cooling air source and the outlet channel configuration is connected with a cooling air sink. A thermal insulation adapted to the collecting basin and the shielding ring as well as to the dual cooling system is preferably put together of austenitic all-metal cassettes. In accordance with an additional feature of the invention, there is provided a further air cooling system in addition to the exterior dual cooling system which is advantageously used for ventilating an upper air cooling chamber, that is disposed above the collecting basin and is limited at its inner periphery by the thermal insulation enclosing the reactor pressure vessel with an annular gap. In accordance with yet another feature of the invention, the collecting basin is penetrated in the upper half of its jacket wall by at least one melt cooling tube which, with a multi-layer construction of the collecting basin, extends through its crucible wall, protective layer, sacrificial material deposit and thermal insulation, is sealed on its inner end by means of a melting plug, extends with a gradient from the outside to the inside and is attached on the inlet side to a cooling liquid reservoir, so that with a core melt present in the collecting basin, the melting plug is heated to its melting temperature and caused to melt and in this way a flow channel to the surface of the core melt for cooling liquid is opened. These features substantially assist in meeting the requirement made under a., above, as well as the requirement c., because surface cooling of the core melt can be achieved thereby. Such surface cooling is not problematic from a safety viewpoint, because the steam, which is not generated suddenly but instead develops continuously, can escape upward through the gaps and cooling gaps that are present and can also condense on the containment walls and the additionally installed recooling heat exchanger heating surfaces, so that the condensed water can again flow into the cooling water reservoir (sump water). Advantageously, the inlet of the melt cooling tube is located outside the supporting and protective structure and is connected with the cooling water reservoir, in which case the melt cooling tube therefore penetrates the circumferential wall of the supporting and protective structure and the spacing gap of the exterior cooling system. With the objects of the invention in view, there is also provided a method for starting and maintaining exterior cooling of a core catcher device of a nuclear reactor installation, which comprises maintaining a cooling water level of the cooling water reservoir at a low water level, during normal operation of the nuclear reactor installation, at which no cooling water can reach the inlet channel configuration of the collecting basin cooling system; feeding emergency cooling water, when a leak occurs in the primary circuit, from the pressure reservoirs to be activated as a function of pressure of the primary circuit into the main coolant lines of the reactor pressure vessel, by feeding the emergency cooling water through the leak location and, if necessary, parallel thereto through further feed locations into the cooling water reservoir; and maintaining a sufficient water volume in the pressure reservoirs to lift the cooling water level of the cooling water reservoir up to a high water level for causing cooling water from the cooling water reservoir to reach the inlet channel configuration and from there the spacing gap of the collecting basin cooling system for filling the cooling system up to the level of the outlet channel configuration, for starting a naturally circulating flow, when the collecting basin is hot, from the cooling water reservoir through the inlet channel configuration to the cooling channels at the bottom wall and the jacket wall of the cooling system and from there through the outlet channel configuration back to the cooling water reservoir. Through the use of this method, according to the object of the invention, the possibility is attained in the case of design or a layout incident to take prepared steps for initiating the naturally circulating cooling of the collecting basin and to employ them. 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 nuclear reactor installation with a core catcher device and a method for exterior cooling of the latter by natural circulation, 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.
	description	This U.S. non-provisional application claims the benefit of priority under 35 U.S.C. § 119 to Korean Patent Application No. 10-2019-0103855, filed on Aug. 23, 2019 in the Korean Intellectual Property Office (KIPO), the contents of which are herein incorporated by reference in their entirety. Various example embodiments relate to a target debris collection device, an extreme ultraviolet light source apparatus including the same, an extreme ultraviolet light system, and/or a method for operating the extreme ultraviolet light source apparatus. More particularly, example embodiments relate to a target debris collection device capable of collecting target material debris formed in an extreme ultraviolet (EUV) vessel and an extreme ultraviolet light source apparatus including the same, etc. In the manufacture of a semiconductor device, extreme ultraviolet (EUV) light may be adopted as a light source in EUV lithography. A laser produced plasma (LPP) source may irradiate a target source having one or more emission rays in the EUV range, e.g., tin, xenon or lithium, with a laser light to emit EUV light by an interaction with the target source and the laser light. When the target source is irradiated by the laser light, microparticles may be formed and deposited within an EUV vessel, and maintenance for collecting and removing the deposited target material debris may be performed in a manual manner. The maintenance of the EUV vessel may take a considerable amount of time, and thus productivity may be decreased. One or more example embodiments provide a target debris collection device capable of automatically collecting and removing target material debris formed within an EUV vessel. One or more example embodiments provide an extreme ultraviolet light source apparatus including the same. According to at least one example embodiment, a target debris collection device for extreme ultraviolet light source apparatus, includes a baffle body extending within an extreme ultraviolet (EUV) vessel between a collector and an outlet port of the EUV vessel, the baffle body configured to allow extreme ultraviolet light reflected from the collector to pass through an internal transmissive region of the baffle body, a discharge plate provided in a first end portion of the baffle body adjacent to the collector, the discharge plate configured to collect target material debris on an inner surface of the baffle body, a guide structure configured to guide the target material debris collected in the discharge plate to a collection tank, and a first heating member provided in the guide structure, the first heating member configured to heat the target material debris. According to at least one example embodiment, an extreme ultraviolet light source apparatus includes a collector included in a vessel, the collector configured to reflect extreme ultraviolet light, a baffle assembly included in the vessel, the baffle assembly configured to allow the extreme ultraviolet light reflected from the collector to pass through an internal transmissive region of the baffle assembly, a guide structure configured to guide target material debris collected in the baffle assembly to a collection tank, and a first heating member provided in the guide structure, the first heating member configured to heat the target material debris. According to at least one example embodiment, an extreme ultraviolet light source apparatus includes a collector included in a vessel, the collector configured to reflect extreme ultraviolet light, and a target debris collection device configured to collect target material debris within the vessel. The target debris collection device includes a baffle body between the collector and an outlet port of the vessel, the baffle body configured to allow the extreme ultraviolet light reflected from the collector to pass through an internal transmissive region of the baffle body, a discharge plate provided in a first end portion of the baffle body adjacent to the collector, the discharge plate configured to collect the target material debris within the baffle body, a guide structure configured to guide the target material debris collected in the discharge plate to a collection tank, the guide structure having a length between 100 mm and 300 mm, and a first heating member provided in the guide structure, the first heating member configured to heat the target material debris. According to at least one example embodiment, a target debris collection device for an extreme ultraviolet light source apparatus may include a guide unit to guide target material debris collected in a baffle assembly to a collection tank. The guide unit may include at least one heating member configured to maintain a liquid and/or semi-liquid state of the target material debris, and/or impede, decrease the occurrence of, and/or prevent the liquid target material debris from becoming solidified. Accordingly, according to one or more of the example embodiments, the target material debris deposited within an EUV vessel may not be solidified (e.g., the target material debris may be in a liquid and/or semi-liquid state), and therefore may be exhausted smoothly to the collection tank. Thus, maintenance time for removing the tin material deposited within the EUV vessel may be reduced greatly. Hereinafter, various example embodiments will be explained in detail with reference to the accompanying drawings. FIG. 1 is a cross-sectional view illustrating an exposure apparatus in accordance with at least one example embodiment. Referring to FIG. 1, an exposure apparatus 10 may include an extreme ultraviolet (EUV) light source apparatus 20, a mirror system 30, a mask stage 40 and/or a wafer stage 50, but is not limited thereto. The mirror system 30 may include an illumination mirror system 32 and/or a projection mirror system 34, etc. In at least one example embodiment, the exposure apparatus 10 may perform a reflective photolithography process using a photomask M. For example, the exposure apparatus 10 may perform an extreme ultraviolet (EUV) exposure process, but is not limited thereto. In particular, the extreme ultraviolet light source apparatus 20 may generate extreme ultraviolet (EUV) light, etc. For example, the extreme ultraviolet light source apparatus 20 may generate light having a wavelength of around 13.5 nm, for example, the wavelength of EUV light, using plasma, etc. A laser generator 120 may direct a laser light L to a target droplet to create a highly ionized plasma P, and the EUV radiation emitted from the plasma may be irradiated to the illumination mirror system of the mirror system 30 through a light collector 210. The illumination mirror system 32 may include a plurality of illumination mirrors, but is not limited thereto. The illumination mirrors may condense the EUV light in order to decrease and/or reduce the loss of the EUV light which propagates out of the mirrored irradiating paths. The mask stage 40 may mount the photomask M on a lower surface thereof and may move in a horizontal direction, but the example embodiments are not limited thereto. For example, the mask stage 40 may move in other directions and/or may mount the photomask M on other surfaces, etc. The photomask M may be mounted on a surface of the mask stage 40 such that a surface (of a front side) in which optical patterns of the photomask M are formed, may face in the direction of a projection mirror system 34, a semiconductor substrate, such as wafer W, etc. The EUV light transferred from the illumination mirror system 32 may be irradiated to the photomask M mounted on the mask stage 40. The EUV light reflected from the photomask M mounted on the mask stage 40 may be transferred to the projection mirror system 34. The projection mirror system 34 may receive the EUV light reflected from the photomask M and transfer the received EUV light to a wafer W (e.g., semiconductor wafer, semiconductor substrate, silicon substrate, etc.). The projection mirror system 34 may include a plurality of projection mirrors. The wafer stage 50 may receive the wafer W and move in a horizontal direction, but is not limited thereto, and the wafer stage 50 may move in additional directions. For example, a photoresist layer having a desired and/or predetermined thickness may be formed on the wafer W, and the EUV light may be focused on the photoresist layer, etc. Accordingly, the exposure apparatus 10 may generate and emit the light onto the photoresist layer on the wafer W, thereby irradiating the photoresist layer. Thus, the photoresist layer may be partially and/or fully exposed based on the optical pattern information of the photomask M to form a photoresist pattern, and then a layer underlying the photoresist pattern may be partially and/or fully etched to form a pattern on the wafer W, doped in accordance with the photoresist pattern, and/or metallized in accordance with the photoresist pattern, etc. Hereinafter, the extreme ultraviolet light source apparatus of the exposure apparatus in FIG. 1 will be explained. FIG. 2 is a block diagram illustrating the extreme ultraviolet light source apparatus in accordance with at least one example embodiment. Referring to FIG. 2, the extreme ultraviolet light source apparatus 20 may include a collector 210 included in an EUV vessel 200 configured to collect and reflect extreme ultraviolet light, a target droplet generator 100 configured to generate and deliver target droplets into the collector 210, and/or a target debris collection device configured to collect target material debris within the EUV vessel 200, but the example embodiments are not limited thereto and may include a greater or lesser number of constituent elements. For example, the extreme ultraviolet light source apparatus 20 may further include a laser generator 120 configured to direct a laser light L to the target droplet injected into the collector 210, etc. Further, the extreme ultraviolet light source apparatus 20 may further include a purge gas supply portion configured to supply a purge gas into the collector 210 as well. In at least one example embodiment, the target droplet generator 100 may generate source droplets 110A, 110B as a target source for generating extreme ultraviolet light. The target droplet generator 100 may inject the source droplets at a desired and/or predetermined period into the collector 210 through a nozzle 102. The target droplet generator 110 may deliver the droplets 110A, 110B of a target material into the interior of the EUV vessel 200 to an irradiation site, that is, a primary focus PF (e.g., a target, etc.) of the collector 210. For example, the target droplet may include at least one element, e.g., xenon, lithium tin, etc., with one or more emission rays in the EUV wavelength range. The EUV emitting element may be in the form of liquid droplets and/or solid particles contained within the liquid droplets, etc. For example, the element tin may be used as pure tin, as a tin compound, e.g., SnBr4, SnBr2, SnH4, as a tin alloy, e.g., tin-gallium alloys, tin-indium alloys, tin-indium-gallium alloys, or combinations thereof. The laser generator 120 may irradiate the target droplet with the laser light L at the irradiation site in the EUV vessel 200. The laser generator 120 may direct the laser light L to the target droplet injected into the collector 210 to generate EUV light. The laser generator 120 may generate a CO2 laser light and then the generated laser light may be focused to the irradiation site through a steering system. The laser light may react with, and vaporize, the target droplet to produce plasma P. The resulting plasma P may emit output radiation, e.g., EUV radiation and/or EUV light. A target material catcher 104 may be installed to be opposite to the nozzle 102 of the target droplet generator 100, but is not limited thereto. The unused or un-irradiated droplets 110C may be collected in the target material catcher 104. The collector 210 may include a reflection surface having the primary focus within or near the irradiation site to which the laser light is focused. According to some example embodiments the reflection surface may be an elliptical reflection surface, but is not limited thereto. The collector 210 may include an aperture 212. The aperture 212 may allow the laser light L to pass through to the irradiation site. The collector 210 may collect, reflect, and/or focus the EUV light (EL) to an intermediate focus IF, and then, the EUV light may be delivered to the mirror system 30 of the exposure apparatus. The purge gas supply portion 130 may supply the purge gas, such as hydrogen gas, etc., into the EUV vessel 200 to transform the target material debris, such as tin, etc., deposited on the inner surfaces of the EUV vessel 200 and/or the collector 210 into volatile target material compounds, such as tin compound SnH4, etc., to thereby purge the volatile target debris compounds (e.g., volatile tin compounds, etc.) from the EUV vessel 200. The target debris collection device may include a baffle assembly 300 in the EUV vessel 200, and the target debris collection device may collect and/or exhaust the target material debris from the EUV vessel 200, e.g., microdroplets formed on the baffle assembly 300, etc. The baffle assembly 300 may include a series of passages and structures that receive, slow, and/or capture a portion of microparticles created when the target droplet is irradiated in the irradiation site. The baffle assembly may extend within the EUV vessel 200 from the collector 210 to an outlet port 202 of the EUV vessel 200, but is not limited thereto. The baffle assembly 300 may not impede, prevent, and/or otherwise occlude the EUV light (EL) from passing from the collector 210 through a three dimensional, cone-shaped transmissive region 302 to the intermediate focus IF. Hereinafter, the target debris collection device will be explained. FIG. 3 is a cross-sectional view illustrating a target debris collection device in accordance with at least one example embodiment. FIG. 4 is a block diagram illustrating a heating device of the target debris collection device in FIG. 3 according to at least one example embodiment. FIG. 5 is a perspective view illustrating a portion of a baffle assembly of the target debris collection device in FIG. 3 according to at least one example embodiment. FIG. 6 is a perspective view illustrating a portion of a discharge plate of the baffle assembly in FIG. 5 according to at least one example embodiment. FIG. 7 is a bottom view of the baffle assembly in FIG. 5 according to at least one example embodiment. FIG. 8 is a perspective view illustrating a target debris collection device in accordance with at least one example embodiment. FIG. 9 is a perspective view illustrating a first heating member of the target debris collection device in FIG. 8 according to at least one example embodiment. FIG. 10 is a perspective view illustrating target material debris discharged along the first heating member of the target debris collection device in FIG. 8 according to at least one example embodiment. Referring to FIGS. 3 to 10, a target debris collection device may include a baffle assembly 300 arranged within and/or included in a EUV vessel 200, and/or a guide unit configured to guide target material debris 112 collected in the baffle assembly 300 to a collection tank 500, but is not limited thereto. The baffle assembly 300 may include a baffle body 310, and/or a discharge plate 320, etc. The guide unit may include at least one heating member configured to impede and/or prevent the liquid target material debris 112 from becoming solidified, or in other words the at least one heating member heats the liquid target material debris 112 in order to decrease the occurrence of the liquid target material debris 112 from becoming solidified, etc. In at least one example embodiment, the baffle body 310 may include a cone-shaped tube extending within the EUV vessel 200 from the collector 210 to the outlet port 202 of the EUV vessel 200, but is not limited thereto. The baffle body 310 may include a first end portion 311a adjacent to the collector 210 and a second end portion 311b opposite to the first end portion 311a and adjacent to the outlet portion 202, but the example embodiments are not limited thereto and may include a single end portion or three or more end portions, etc. The first end portion 311a may have a diameter greater than a diameter of the second end portion 311b, but is not limited thereto, and the first end portion 311a may have the same diameter as the second end portion 311b, etc. The extreme ultraviolet layer EL reflected from the collector 210 may be directed to the intermediate focus IF through the cone-shaped transmissive region 302 of the baffle body 310. A plurality of vanes 312 may extend in an extending direction of the baffle body 310 on an inner wall of the baffle body 310. The vane 312 may protrude in a radial direction inwardly from the inner wall of the cone-shaped baffle body 310. The vanes 312 may surround the transmissive region 302, and may not protrude into the transmissive region 152, but is not limited thereto. The baffle body 310 may be configured in a near vertical orientation, but is not limited thereto. A central line CL of the baffle body 310 may be oriented to be inclined at a desired and/or predetermined angle θ with respect to a gravity direction G. For example, the angle θ may be in a range of from 1 degree to 30 degrees, but is not limited thereto. Additionally, as described later, the baffle body 310 may be heated by at least one other heating member, such as a fourth heating member 360. For example, the vanes 312 of the baffle body 310 may be heated to a temperature of about 100° C. to about 400° C., but the example embodiments are not limited thereto, and for example, the vanes 312 may be heated to a temperature appropriate to maintain the target material debris in liquid or semi-liquid form. The discharge plate 320 may be provided in the first end portion 311a of the baffle body 310, but is not limited thereto. The discharge plate 320 may include a plate having an annular shape, but is not limited thereto. The vanes 312 may extend from an upper surface of the discharge plate 320 along the extending direction of the baffle body 310, but is not limited thereto. The target material debris may be collected by the discharge plate 320 which is positioned at a relatively low level, e.g., the discharge plate 320 towards a location away from the outlet port 202 of the EUV vessel 200, etc., and the target material debris collected on the discharge plate 320 may be exhausted from the baffle assembly through a single passage or a series of passages. For example, first and second discharge passages 322, 324 and a connection passage 326 connecting the first and second discharge passages may be formed in the upper surface of the discharge plate 320, and a discharge hole 340 connected thereto may be formed in the discharge plate 320, but the example embodiments are not limited thereto. The first and second discharge passages 322, 324 may be formed concentrically around the center of the discharge plate 320, but are not limited thereto. At least two discharge holes 340 may be formed in the discharge plate 320, but the example embodiments are not limited thereto, and for example, a single discharge hole may be present, etc. A discharge nozzle 350 may be installed in a lower surface of the discharge plate 320 to be connected to the discharge hole 340. The discharge nozzle 350 may protrude from the lower surface of the discharge plate 320. For example, the discharge nozzle 350 may have a diameter (D) between approximately 4 mm to 16 mm, but is not limited thereto. In at least one example embodiment, a guide structure may be connected to the discharge nozzle 350 such that the target material debris 112 collected on the discharge plate 320 is guided to be discharged to the collection tank 500. Additionally, at least one first heating member may be provided in the guide structure to heat the liquid target material debris 112, and therefore impede, decrease the possibility of, and/or prevent the liquid target material debris 112 from becoming solidified. As illustrated in FIGS. 6 to 10, the guide structure 400 may extend from the discharge plate 320 toward the collection tank 500. The guide structure 400 may include a plurality of heating lines 410 as the first heating member. The heating line 410 may extend in a direction parallel with the gravity direction from the discharge nozzle 350 toward the collection tank 500, but is not limited thereto. The heating line 410 may be the first heating member connected to a first power supply 450. Additionally, the heating line 410 may serve as at least a portion of the guide structure 400. In this case, an outer covering material (e.g., insulating or conductive material) of the heat line may serve as the guide structure 400. For example, three heating lines 410 may be provided. The three heating lines 410 may be spaced apart from one another. In this case, the liquid target material debris 112 may run down between the three heating lines 410 to be collected in the collection tank 500. The guide structure may have a length of approximately 100 mm to approximately 300 mm, but the example embodiments are not limited thereto. For example, the length of the guide structure may be determined in consideration of a distance between the discharge nozzle 350 and the collection tank 500, etc. The first heating member 410 may be electrically connected to the first power supply 450. The liquid target material debris 112 may be heated by the first heating member 410 such that the liquid target material debris is collected in the collection tank 500 without becoming solidified. As illustrated in FIG. 4, a second heating member 420 may be provided in the discharge nozzle 350 to heat the target material debris 112 to maintain the target material debris in a liquid or semi-liquid state, and therefore impede, decrease the occurrence of, and/or prevent the target material debris from becoming solidified. The second heating member 420 may include a heating line which surrounds the discharge nozzle 350. The second heating member 420 may be electrically connected to a second power supply 452. A third heating member 430 may be provided in the passage of the discharge plate 320 to heat the target material debris 112 to maintain the target material debris in a liquid or semi-liquid state, and therefore impede, decrease the occurrence of, and/or prevent the target material debris from becoming solidified. The third heating member 430 may include a heating line which extends along the first and second discharge passages 322, 324 and the connection passage 326. The heating line may extend from the passage to the discharge hole 340. The third heating member 430 may be electrically connected to a third power supply 454. The fourth heating member 360 may be provided in the baffle body 310 to heat the target material debris 112 to maintain the target material debris in a liquid or semi-liquid state, and therefore impede, decrease the occurrence of, and/or prevent the target material debris from being solidified. The fourth heating member 360 may include a heating line which surrounds the baffle body 310. The fourth heating member 360 may be electrically connected to a fourth power supply 352. The first to fourth power supplies 450, 452, 454, 362 may be connected to a controller. The controller may control temperatures of the first to fourth heating members 410, 420, 430, 360 to heat the target material debris 112 to maintain the target material debris in a liquid or semi-liquid state, and therefore impede, decrease the occurrence of, and/or prevent the target material debris from becoming solidified. As mentioned above, the target debris collection device may include the guide unit to guide target material debris 112 collected in the baffle assembly 300 to the collection tank 500. The guide unit may include the at least one heating member configured to heat the target material debris 112 to maintain the target material debris in a liquid or semi-liquid state, and therefore impede, decrease the occurrence of, and/or prevent the liquid target material debris 112 from being solidified. Accordingly, the target material debris deposited on the EUV vessel 200 may be maintained in a liquid or semi-liquid state (e.g., not become solidified), and may be exhausted smoothly to the collection tank 500. Thus, maintenance time for removing the target debris material (e.g., tin material, etc.) deposited in the EUV vessel may be greatly reduced. FIG. 11 is a perspective view illustrating a target debris collection device in accordance with at least one example embodiment. The target debris collection device may be substantially the same as or similar to the target debris collection device described with reference to FIGS. 3 to 10 except for a configuration of a guide unit. Thus, same reference numerals will be used to refer to the same or like elements and any further repetitive explanation concerning the above elements will be omitted. Referring to FIG. 11, a target debris collection device may include a baffle assembly 300 arranged within a EUV vessel 200, a guide structure 400 configured to guide target material debris 112 collected in the baffle assembly 300 to a collection tank 500, and/or a first heating member 410 provided in the guide structure 400 to heat the target material debris 112 to maintain the target material debris in a liquid or semi-liquid state, and therefore impede, decrease the occurrence of, and/or prevent the liquid target material debris 112 from becoming solidified, etc. In at least one example embodiment, the guide structure 400 may include a guide plate which extends from a discharge plate 320 toward the collection tank 500. For example, the guide plate may have a cross-section of circle or semi-circle, but is not limited thereto. The guide structure 400 may extend in a direction parallel with a gravity direction from a discharge nozzle 350 toward the collection tank 500. The first heating member 410 may include a heating line which extends along the guide plate. The first heating member 410 may be electrically connected to a first power supply 450. The guide plate may protect liquid target material debris 112 from deviating by a gas flowing within the EUV vessel 200 and may guide the liquid target material debris 112 to the collection tank 500. Accordingly, the guide plate may protect the liquid target material debris from being affected by the hydrogen gas current such that the liquid target material debris may be collected precisely into the collection tank 500. The above extreme ultraviolet light source apparatus may be applied in applications such as a lithography apparatus for manufacturing semiconductor devices, however, the example embodiments are not limited thereto. The extreme ultraviolet light source apparatus according to at least one example embodiment may be applied to other photolithography processes, such as processes for manufacturing display devices such flat display, organic light emitting display, etc. The foregoing is illustrative of various example embodiments and is not to be construed as limiting thereof. Although a few example embodiments have been described, those skilled in the art will readily appreciate that many modifications are possible in the example embodiments without materially departing from the novel teachings and advantages of the inventive concepts. Accordingly, all such modifications are intended to be included within the scope of the example embodiments as defined in the claims.
	abstract	Systems and methods are provided for reducing the storage time of spent nuclear fuel. In one embodiment, a method is provided that includes providing a sample of spent nuclear fuel and irradiating the spent nuclear fuel with substantially collimated gamma ray photons having energy levels of about 10 MeV to about 15 MeV for a predetermined time period to initiate a photofission reaction in the remaining fertile fissile material in the spent nuclear fuel.
062563637	summary	FIELD OF THE INVENTION The present invention relates to a storage/transport container for spent nuclear-fuel elements. More particularly this invention concerns such a container used for spent fuel rods. BACKGROUND OF THE INVENTION A transport/storage container for spent nuclear-fuel elements typically has a vessel having a side wall with an inner surface defining an interior extending along an axis and a multilevel basket extending substantially a full axial length of the interior and forming a plurality of axial full-length rectangular-section wells adapted to receive the spent fuel elements. The lower end of the interior is closed by a permanent floor and the upper end by a massive but removable cover. It is essential to maintain the rods held in such a container in a subcritical state. Thus their neutron emissions must be controlled. This is normally accomplished as described in U.S. Pat. No. 5,032,348 by forming each level of the basket of a plurality of fitted-together plates. Each such set of plates forms a plurality of openings that together form the wells that receive the spent fuel rods. All the plates fit together at slot joints at the corners of all the wells. The neutron-absorbing plates are normally of a relative poor thermal capacity so it is necessary to alternate layers made of plates of neutron-absorbing material with layers made of plates of a more thermally conductive material so that the heat generated in such a container can be conducted to the side walls. Such a system therefore trades dissipating heat off against suppressing neutron emissions. One function can only be made better by making the other worse. In addition fitting together the numerous plates making up each level is an onerous task, involving meticulously fitting together long plates simultaneously at multiple joints so that the container is expensive to manufacture of a large number of pieces of different sizes. In another system borated-steel plates form the baskets. Plates of this material must be welded or screwed together, as the slot joint described above is not usable because of the brittleness of borated steel. Thus assembly of the basket becomes extremely expensive since it is too complex to automate, even with welding which is the preferred and cheaper solution. Furthermore the container is extremely heavy when its basket is made of steel. It is further known for some of the wells in a borated-steel basket to be water filled, acting as neutron traps. Such use of water makes the container quite large and of course also increases its weight, while making the gaps necessary for flow of the water increases the cost of the plates. Furthermore using some of the wells for water only reduces the number of wells usable for fuel rods, reducing the capacity of the container. The basket furthermore is not particularly strong and has difficulty meeting requirements in this regard, as the basket must keep the fuel rods apart even if dropped or otherwise subjected to some serious axial and/or radial stress. Finally, such a system is difficult to decontaminate. OBJECTS OF THE INVENTION It is therefore an object of the present invention to provide an improved container for spent nuclear-fuel rods. Another object is the provision of such an improved container for spent nuclear-fuel rods which overcomes the above-given disadvantages, that is which is of inexpensive and light construction and considerable neutron-absorbing capacity yet which has a high load capacity and is easy to decontaminate. SUMMARY OF THE INVENTION A transport/storage container for spent nuclear-fuel elements has according to the invention a vessel having a side wall with an inner surface defining an interior extending along an axis and a plurality of like basket sections forming a stack extending substantially a full axial length of the interior and forming a plurality of axial full-length rectangular-section wells adapted to receive the spent fuel elements. Each of the basket sections is formed of two long light-metal neutron-absorbing plates crossing each other, each having a pair of outer ends directly engaging the inner surface of the side wall in heat-transmitting contact therewith, and subdividing the interior at the respective section into a plurality of segments. A plurality of short light-metal neutron absorbing plates are fitted together in each of the segments and form with the main plates of the respective section rectangular-section axially throughgoing openings forming the wells with the plates of the other sections. The short plates are not actually connected to the long plates. Thus these long plates can fulfill the heat-conducting function that is preferred at each level. In addition the use of interfitted short plates in each segment of each level means that these short plates are easily assembled together. In a standard system with eight wells per segment there will be at most two joints between any short plate and the other plates. Such a system is still as strong as needed, even though the short plates are not actually joined to the long plates. The wells are normally defined on at least two sides by the plates and, toward the center of the container they are defined on all four sides by the plates. For maximum strength and lightness the plates are made of a borated light metal, preferably aluminum or an aluminum alloy. Normally the plates are all of the same material, since the long plates will also work effectively to conduct heat to the wall of the container. Preferably according to the invention the long plates are thicker than the short plates for best heat conduction and strength. More particularly the long plates are between 9 mm and 15 mm, preferably 11 mm and 13 mm, thick. The short plates are between 6 mm and 10 mm, preferably 7 mm and 9 mm, thick. With a system having 32 wells, only two different lengths of short plates are needed, plates having a length equal to twice the width of a well, and plates with three times that width. In accordance with the invention the inner surface is generally cylindrical and centered on the axis. The long plates are perpendicular to each other and subdividing the interior at the respective level into four quadrants. In addition at least one respective filler block is provided in each of the quadrants. The blocks each have a curved outer surface complementarily engaging the inner surface at the respective level and at least one planar inner surface forming walls of respective openings. Furthermore each filler block is provided with shielding. These blocks can be made of cast borated aluminum with lead bodies imbedded in them. Furthermore the plates each have upper and lower edges which, except for at the bottom and top levels, engage lower and upper edges of the overlying and underlying plates. The wells therefore basically continuous side walls so that any neutrons will be intercepted. Ends of the short plates bear in heat-transmitting relationship on the main plates and on the filler blocks so that, even though these short plates are normally somewhat thinner than the long plates, they also transmit heat effectively. There is no actual connection of the short plates to the long plates, for instance by means of half-width joint slots; at most the short plates bear longitudinally on the main plates. Thus these filler blocks, which have the same axial dimension as the plates, have three functions: conducting heat away from the contained rods to the container walls, blocking neutrons, and transversely bracing the basket sections and tubes in the container. According to the invention a respective light-metal tube extends substantially the full length of each of the wells. The spent nuclear-fuel elements are within the light-metal tubes. These tubes are each provided with axial guide passages and is provided therein with neutron-absorbing rods. Each tube is of the same square or rectangular section as the respective well and fits tightly therein so that these tubes considerably rigidify the system, holding the basket sections in accurate axial alignment with one another. These tubes can be of stainless steel. The long plates are in accordance with the invention formed with interfitting half-width joint slots. The short plates are similarly joined to one another, but not to the main plates. The elimination of welding considerably reduces the cost of manufacturing the container. The system is thus normally built level-by-level, fitting all the plates and filler blocks for one level together in the container before starting the next level. This modular construction allows the same basic parts to be used in containers of different axial dimensions. Once all the basket sections are in place, the stainless-steel tubes are installed to lock everything together. Such a container can be quite strong and still relatively light. It is cheap and easy to assembly, and none of its parts require complex manufacture or machining. The combination of the stiffening tubes and the basket sections together produces an extremely rigid assembly. The container of this invention has a higher load capacity, that is the ability to hold more fuel rods, than the above-discussed prior-art systems. It is possible to eliminate the need for water passages between adjacent basket sections with the highly efficient and shielded system of this invention. The elimination of borated steel greatly reduces the cost and weight of the system. Only three different plates are needed, one type of long plate and two lengths of short plate, so construction costs can be reduced considerably. Similarly two types of filler blocks are all that are needed, ones with one flat side that closes two adjacent wells and another with two separate flat sides that close sides of two adjacent wells.
051241153	summary
	abstract	A method of carrying equipment out of a nuclear reactor building that, in carrying equipment rendered radioactive out of a nuclear reactor building, makes it possible to shorten the operating time and reduce the number of casks to be used, wherein the structures of a nuclear power plant that are outside the equipment installed in the nuclear power plant are removed and these removed structures are then stored in the space in the equipment and carried out of the building.
047643358	summary	BACKGROUND OF THE INVENTION This invention pertains in general to a system for detecting the presence of failed fuel elements within the core of a nuclear reactor and more particularly, to an on-line system for diagnosing the identity and condition of breached fuel elements in a nuclear reactor. The reactor fuel in a fission type reactor is typically an isotope of uranium, such as uranium 235. The reactor fuel may take the form of a fluid, such as an aqueous solution of enriched uranium; but typically the fuel is solid, either metallic uranium or a ceramic such as uranium oxide or uranium-plutonium oxide. The solid fuel material is fabricated into various small plates, pellets, pins, etc; which are usually clustered together in an assemblage called a fuel element. Almost all solid fuel elements are clad with a protective coating or sheath that prevents direct contact between the fuel material and the reactor coolant. The cladding also serves as part of the structure of the fuel elements. The operation of the fuel elements generates heat, which heat is typically dissipated by means of a coolant passed through the reactor. The coolant can be water operating as either liquid or steam, or the coolant can be a liquid metal such as sodium or a sodium-potassium mixture. The coolant passes in proximate contact over the cladded-fuel elements; and sound cladding isolates or separates the coolant from the radioactive fuel material. However, in the event of a breach in the cladding, the coolant directly contacts the fuel. The radioactive discharge may then in turn be conveyed via the coolant throughout the entire coolant system thereby contaminating the entire system. Also given off, as part of the radioactive discharge, are at least nine different isotopes that not only give off typical gamma rays of radioactivity, but also give off what are known as delayed neutrons. These isotopes, or delayed neutron emitters, would include bromine, iodine, and tellurium to name a few. Each of these delayed neutron emitters is soluble in liquid sodium (the coolant) so that it readily blends in with the coolant, should a fuel element cladding breach occur, and flows from the coolant throughout the system. Therefore, it becomes readily apparent that the event of fuel cladding breaches must be taken into account when designing and operating a nuclear reactor. A quick and precise diagnosis of cladding breach events would ensure that the reactor operator would correctly respond upon the occurence of such a breach. A precise diagnosis of the condition of a breached pin would introduce significant advantages for reactor plant operation. The reactor could be safely operated under such breached pin conditions until a predetermined allowable radioactivity limit is exceeded. In the event that this predetermined limit is not exceeded, the reactor could be safely operated until the next scheduled discharge. Further, if a system could accurately diagnose whether a cladding breach is stable or unstable, the reactor operator could continue the operation of the reactor in the event of a stable breach and shut down the reactor upon the occurrence of an unstable breach. The continued operation of the reactor under a stable breached pin condition could significantly improve reactor availability. Conventionally, radioactive elements which have mixed with the coolant are detected by means of a GeLi detector (a germanium and lithium gamma-ray detector) incorporated into a GLASS (a germanium-lithium argon scanning system), or other readily available detecting systems. These systems are used to detect the activity in the cover gas of the reactor. Typically however, these systems have been used only as an annunciation of fuel failure. After identification of a "gas leaker" (a breached fuel element), fission gas activity is typically removed by a plant cleaning system, such as the cover gas cleanup system (CGCS) used in EBR-II. This system removes fission gas activity from the reactor cover gas by semicontinuously extracting part of the cover-gas, cleaning it cryogenically and returning it to the core. The CGCS allows the reactor to be operated after a breach has occurred in a fuel element. However, the system effectively obscures any information about the failure which may be contained in the gas release data obtained from a GLASS. Therefore, typically used in conjunction with CGCS are delayed neutron systems such as the system disclosed in U.S. Pat. No. 4,415,524 issued to K. C. Gross et al., and/or fuel element failure location systems, such as the gas tagging system disclosed in U.S. Pat. No. 4,495,143, issued to K. C. Gross et al., which monitor the identity and condition of breached pins. However, such systems provide only very qualitative information about the type of breached fuel involved (oxide or metal), and its burnup (high or low). The run beyond clad breach mode operation of a commercial liquid metal reactor may not be allowed without an on-line identification of breached pins and a diagnosis of the breached pin condition and development. The diagnosis of a breached pin should include a reliable prediction of the on-going condition of the event. Therefore, in view of the above, it is an object of the present invention to provide an on-line apparatus and method for diagnosing the severity of a breached fuel element. It is another object of the present invention to provide an on-line apparatus and method for determining the number of breached fuel elements. It is another object of the present invention to provide an apparatus and method for determining the mode of gas released from a breached fuel element. It is a further object of the present invention to provide an apparatus and method for determining the breaching mechanism in a fuel element. It is still a further object of the present invention to provide an apparatus and method for determining if a breach in the cladding of a fuel element is benign or unstable. Additional objects, advantages and novel features of the invention will be set forth in part in the description which follows, and in part will become apparent to those skilled in the art upon examination of the following or may be learned by practice of the invention. The objects and advantages of the invention may be realized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims. SUMMARY OF THE INVENTION To achieve the foregoing, and other objects the present invention, the present disclosure provides an apparatus and method for diagnosing breached fuel elements in a nuclear reactor having a cover-gas cleanup system. According to the present invention, a detection system measures the activity from isotopes in the cover gas of a reactor. A data acquisition and processing system monitors the output of the detection system and corrects for the effects of the cover gas cleanup system on the measured gas activity. The data acquisition and processing system further calculates the curves of the derivative of the corrected gas activity as a function of time for each measured isotope. A display means exhibits graphs the curves of the corrected gas activity and derivative thereof as a function of time. The derivative curve represents the instantaneous release rate of fission gas from a breached fuel element.
054835658	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS FIGS. 1-4 show a fuel channel 1 with a substantially square cross section. The fuel channel 1 surrounds with no significant play an upper square portion 2a of a transition section 2 which otherwise comprises a conical portion 2b and a cylindrical portion 2c (see also FIGS. 5a-c). The transition section 2 has a downwardly-facing inlet opening 3 for cooling water. The transition section 2 is formed with a guide member which is intended to guide the fuel assembly 16 into an assembly supporting plate (not shown) and comprises a plurality of guiding spokes 2d. Besides supporting the fuel channel 1, the transition section 2 also supports a bottom support 4. At its bottom the fuel channel 1 has a relatively thick wall portion which is fixed to the transition section 2 and the bottom support 4 by means of a plurality of horizontal bolts 5. According to the embodiment shown, with a hollow support member 6 of cruciform cross section, the fuel channel 1 is divided into four vertical tubular parts 7a-d with at least substantially square cross section. The support member 6 is welded to the four walls 1a-d of the fuel channel 1 and has four hollow wings 8. The central channel formed by the support member 6 is designated 9 and at its bottom extended down through the bottom support 4 with an inlet 10 for moderator water. Each tubular part 7a-d comprises a bundle of twenty-four fuel rods 12. The rods 12 are arranged in a symmetrical lattice in rows in which each rod 12 is included in two rows perpendicular to each other. Each bundle is arranged with a bottom tie plate 13, a top tie plate 14 and a plurality of spacers 15. A fuel rod bundle with a bottom tie plate 3, a top tie plate 14, spacers 15 and fuel channel part 1 forms a unit which is referred to as a sub-assembly, whereas the device illustrated in FIGS. 1-4 and comprising four such sub-assemblies is referred to as a fuel assembly 16. A unit comprising four fuel assemblies 16 and a control rod 17 arranged centrally therebetween constitutes a supercell. The spaces between the fuel rods 12 within each sub-assembly 7 are traversed by water, as is the hollow support member 6 of cruciform cross section in the fuel assembly 16. The gaps 18a-b between the fuel assemblies 16 are also traversed by water. The four bottom tie plates 13 are supported in the fuel assembly 16 by the bottom support 4 and are each partially inserted into a respective square hole therein. The holes for the passage of the water through the bottom tie plate 13 are designated 19. FIG. 3 shows part of a symmetrical core lattice according to the prior art. The section comprises a supercell. In a symmetrical core lattice, the control rod gaps 18a, into which the control rods 17 can be inserted, have the same width as the narrow gaps 18b, into which no control rods 17 can be inserted. The control rods 17 have blades which form a rectilinear cross and are arranged centrally in the supercell. Two continuous holes 20a-b for by-pass flow are arranged in the wall portion of the transition section 2, facing away from the control rod 17. To illustrate the by-pass holes more clearly in the FIGS. 3 and 4a, one fuel assembly 16 in each figure is shown without the cruciform water channel 19. FIGS. 4a-c show part of a symmetrical core lattice, a supercell, according to the invention. FIG. 4a shows a supercell with fuel assemblies with by-pass holes 20a-d arranged in the transition section in a position corresponding to the center of the side surfaces of the fuel channel 1. FIGS. 5a-b show how the transition section is provided with a turnable valve, a turning ring 21, with holes 22a-b arranged such that a turning of the ring allows opening of two holes 21a-d at a time, which are located such that the by-pass flow is directed away from the control rod. The by-pass holes 20a-d may be arranged at any location between the assembly supporting plate and the lowest part of a fuel pellets column 24. FIG. 4b shows a supercell with fuel assemblies where the by-pass holes 20a-d are arranged in a position corresponding to the corner of the fuel channel 1 where the transition section 2, in a manner corresponding to the embodiment of FIG. 4a, is provided with a turning ring 21 according to FIGS. 5a-b. According to FIG. 4a, in the original operating position of the fuel assembly 16, the by-pass flow passes through the two holes 20a and 20b which are facing the narrow gaps 18b, whereas in the operating position when the fuel assembly 16 has been turned, for example, 180.degree., the by-pass flow passes through the holes 20c and 20d which in the original operating position were facing the control rod gaps 18a, but which after the turning are facing the narrow gaps 18b. This change of flow is made possible by causing the turning ring 21 to open/close two of the four holes 20a-d with which the transition section 2 is provided. Thus, by turning the turning ring 21, the two holes 20c and 20d, which in the original operating position are closed and facing the control rod gap 18a, open for the by-pass flow whereas the other two holes 20a and 20b are closed. In FIG. 4b, in the original position of the fuel assembly 16, the by-pass flow passes through the holes 20b and 20d whereas the holes 20a and 20c are closed by means of the turning ring 21. When turning the upper lefthand fuel assembly 16, shown in the figure, for example 90.degree. in the clockwise direction, the by-pass flow, after turning the turning ring 21, instead passes through the holes 20a and 20c when are then facing away from the control rod 17 in the supercell. FIG. 4c shows an embodiment intended for fuel assemblies 16 which can only be turned 180.degree. around their longitudinal axis. The embodiment shows two holes 20a-b for by-pass flow. The holes 20a-b are diametrically opposed and arranged symmetrically in the wall portion of the transition section 2 such that both holes 20a-b substantially have the same distance to the control rod 17. No turning ring 21 is needed since a 180.degree. turn of the fuel assembly 16 results in the location of the holes 20a-b in relation to the control rod 17 being the same as before the turning. A transition section 2 according to FIG. 4c is also clear from FIG. 5c. FIG. 5b shows the design of the turning ring 21 and its arrangement around the transition section 2. The front ring 23 shown in FIG. 5b is intended to fix the turning ring 21 in the desired position. In this case, the transition section 2 is provided with four holes 20a-d for the by-pass flow arranged around the conical wall portion 2b of the transition section 2 with an approximately 90.degree. pitch angle. The turning ring 21 is provided with two holes 22a-b.
	claims	1. An elastomeric material comprising:an elastomer in which is dispersed a powder of metal oxides,wherein the powder of metal oxides comprises from 70 to 90% by mass of bismuth trioxide, from 5 to 15% by mass of tungsten trioxide, and from 5 to 15% by mass of lanthanum trioxide, such that said elastomeric material retains mechanical flexibility,wherein said elastomeric material is constructed so as to attenuate ionizing radiation in accordance with a gamma-radiation attenuation factor, andwherein said elastomer and said powder of metal oxides are selected in varying proportion to provide a particular mechanical flexibility and gamma-radiation attenuation factor. 2. The material according to claim 1, wherein the elastomer represents from 15 to 35% by mass of the mass of the elastomeric material and the powder of metal oxides represents from 65 to 85% by mass of the mass of the elastomeric material. 3. The material according to claim 2, wherein the elastomer represents 25±2% by mass of the mass of the elastomeric material and the powder of metal oxides represents 75±2% by mass of the mass of the elastomeric material. 4. The material according to claim 1, wherein the powder of metal oxides comprises 80±2% by mass of bismuth trioxide, 10±1% by mass of tungsten trioxide, and 10±1% by mass of lanthanum trioxide. 5. The material according to claim 1, wherein the powder of metal oxides consists of particles, at least 90% by number of which have a size of between 1 and 100 μm. 6. The elastomeric material according to claim 1, wherein the powder of metal oxides consists of particles, at least 80% by number of which have a size of between 1 and 50 μm. 7. The material according to claim 1, wherein the elastomer is selected from natural rubber, synthetic polyisoprenes, polybutadienes, polychloroprenes, chlorosulfonated polyethylenes, elastomeric polyurethanes, fluorinated elastomers, isoprene-isobutylene copolymers, ethylene-propylene-diene copolymers, styrene-isoprene-styrene block copolymers, styrene-ethylene-butylene-styrene block copolymers, and mixtures thereof. 8. The material according to claim 7, wherein the elastomer is a polychloroprene. 9. The material according to claim 1, further comprising one or more adjuvants selected from plasticizers, flexibilizing agents, antistatic agents, lubricating agents, adherence promoters, and coloring agents. 10. A protective article for protecting against ionizing radiations, said protective article comprising:an elastomeric material in which is dispersed a powder of metal oxides,wherein the powder of metal oxides comprises from 70 to 90% by mass of bismuth trioxide, from 5 to 15% by mass of tungsten trioxide, and from 5 to 15% by mass of lanthanum trioxide, such that said elastomeric material retains mechanical flexibility,wherein said elastomeric material is constructed so as to attenuate ionizing radiation in accordance with a gamma-radiation attenuation factor, andwherein said elastomeric material and said powder of metal oxides are selected in varying proportion to provide a particular mechanical flexibility and gamma-radiation attenuation factor. 11. The article according to claim 10, wherein said article is an apron, chasuble, jacket, skirt, glove, sleeve, thyroid protection, gonad protection, eye protection band, mammary protective bra, surgical drape, or a curtain. 12. The article according to claim 10, wherein said article is a glove. 13. A protective multilayer glove for protecting against ionizing radiations, comprising:a first layer comprising at least one layer of an elastomeric material inserted between at least two other layers of another elastomeric material,wherein said elastomeric material of said at least first layer comprises an elastomeric material in which is dispersed a powder of metal oxides,wherein the powder of metal oxides comprises from 70 to 90% by mass of bismuth trioxide, from 5 to 15% by mass of tungsten trioxide, and from 5 to 15% by mass of lanthanum trioxide, such that said elastomeric material retains mechanical flexibility,wherein said elastomeric material is constructed so as to attenuate ionizing radiation in accordance with a gamma-radiation attenuation factor, andwherein said elastomeric material and said powder of metal oxides are selected in varying proportion to provide a particular mechanical flexibility and gamma-radiation attenuation factor. 14. The glove according to claim 13, wherein said at least two other layers comprise a second layer and a third layer in an elastomeric material selected from natural rubber, synthetic polyisoprenes, polybutadienes, polychloroprenes, chlorosulfonated polyethylenes, elastomeric polyurethanes, fluorinated elastomers, isoprene-isobutylene copolymers, ethylene-propylene-diene copolymers, styrene-isoprene-styrene block copolymers, styrene-ethylene-butylene-styrene block copolymers, and mixtures thereof. 15. The glove according to claim 14, wherein said second and third layers are in elastomeric polyurethane. 16. The glove according to claim 13, wherein each of the layers of the glove has a thickness from 50 to 1,500 μm. 17. The glove according to claim 16, wherein the first layer has a thickness from 50 to 200 μm and each of the at least two other layers has a thickness from 150 to 300 μm. 18. The glove according to claim 13, wherein the glove comprises a sleeve having a same composition as the glove, having a length of from 25 to 100 cm. 19. The glove according to claim 13, wherein the glove is adapted to protect against ionizing radiations emitted by powders of nuclear fuels. 20. The glove according to claim 13, wherein the at least two other layers are identical. 21. The material according to claim 1, wherein said gamma-radiation attenuation factor is from 1.5 to 4. 22. The material according to claim 1, wherein said mechanical flexibility is defined as an elasticity modulus of from 1.0 to 1.9 Mpa. 23. The article according to claim 10, wherein said gamma-radiation attenuation factor is from 1.5 to 4. 24. The article according to claim 10, wherein said mechanical flexibility is defined as an elasticity modulus of from 1.0 to 1.9 Mpa. 25. The glove according to claim 13, wherein said gamma-radiation attenuation factor is from 1.5 to 4. 26. The glove according to claim 13, wherein said mechanical flexibility is defined as an elasticity modulus of from 1.0 to 1.9 Mpa.
047924292	summary	FIELD OF THE INVENTION This invention relates to nuclear reactor fuel assemblies and in particular those assemblies which include spaced fuel rod support grids mounted in a reactor core as a unit. The fuel rods are held between an upper end fitting or top nozzle and a lower end fitting by means of spacer grids. The reactor coolant flows upwardly from holes in the lower end fitting along the fuel rods, and upwardly through holes in the upper end fitting. When the fuel assembly is loaded in a reactor core, an upper core plate over the fuel assembly reacts against fuel assembly holddown leaf spring assemblies attached by fasteners to the upper end fitting to provide a downward force. This force combines with the fuel assembly weight to prevent fuel assembly liftoff from hydraulic forces during operation of the reactor pumps. BACKGROUND OF THE INVENTION Debris in the circulating coolant which collects or is trapped in fuel rod spacer grids is believed responsible for as many as 30% of known fuel rod failures. Laboratory and in-reactor experience indicate that fuel rod cladding failures can be caused by debris trapped in a grid region which reacts against the fuel rod cladding in a vibratory fashion resulting in rapid wear of the cladding. The size and shape of the debris capable of causing severe damage is quite variable and may include broken fuel assembly fasteners. The invention involves a spring retention cap for a nuclear fuel assembly end fitting leaf spring assembly which is designed to prevent the creation of debris, such as pieces of broken holddown spring retention screws. The cap of the invention also minimizes the loss of holddown force, the chance that a failure of one screw will impart a jacking force which leads to the failure of a second screw, and interference with control rod operation because of rotation of a leaf spring. Moreover, the cap is designed to provide simple installation and removal procedures. The cap also saves fuel reconstitution expense, because it eliminates the requirement of welds for preventing screw rotation. More importantly, the unique cap provides the option of continuing to operate safely, without reconstitution, even if a holddown spring retention screw breaks. SUMMARY OF THE INVENTION The spring retention cap of the invention for a nuclear fuel assembly end fitting or top nozzle having a leaf spring assembly secured by spring retention fasteners is a body with two plane exterior surfaces defining an angled corner. The cap has a base for engaging the end fitting and a leaf spring assembly receiving slot in the base spaced from and extending substantially parallel to the two plane exterior surfaces. An inwardly directed flange is formed at the slot opening by a portion of the base to define a connecting hook-like structure which engages a slot in each of two end fitting exterior surfaces corresponding to the two exterior plane surfaces of the cap. The base has a portion on the opposite side of the slot in the base from the flange which is inward of the hook-like structure defined by the flange. A screw secures the retention cap in position with the ends of the leaf springs and the spring retention screws covered thereby. The distance from the two spaced exterior surfaces of the body to the slot, combined with the cross-section of the screw, are sufficient to provide strength to accommodate a jacking force created by the leaf spring assembly in the event of failure of one of the two leaf spring assembly retention screws. No loading is placed on the other spring retention screw as in some other designs. If the other screw should also fracture, both leaf spring assemblies and retention screws will be captured by the cap and its attachment screw. This will preclude the creation of loose parts. If there are no problems with the spring retention screws, the only function of the cap and the third screw is to protect the spring screws and spring ends from being impacted during fuel handling operations.
	description	The present application claims priority to Japanese Patent Application No. JP2008-192950 filed Jul. 28, 2008, the entire contents of this application being incorporated herein by reference in their entirety. 1. Field of the Invention The present invention relates to apparatuses capable of performing frequency adjustment by etching elements formed on a wafer, or relates to apparatuses capable of performing frequency adjustment by applying a frequency adjusting material to elements formed on a wafer. 2. Description of the Related Art In a conventional frequency adjusting method, the frequency of a piezoelectric element is adjusted by etching the piezoelectric element with an ion beam. To achieve higher productivity in using this method, a plurality of elements are formed or arranged on a wafer so that the frequencies of the plurality of elements can be adjusted at the same time. In this frequency adjusting method, to prevent the wafer other than areas of desired elements from being irradiated with the ion beam, it is necessary that the wafer be masked with a pattern mask. However, when each element is small in size, it is difficult to selectively apply the ion beam only to a desired element. Therefore, a plurality of elements may be grouped together as a single irradiation area to be irradiated. When such small elements are closely arranged on the wafer, it is necessary to apply the ion beam uniformly to the entire surface of the wafer. However, depending on the size or shape of a hole in the pattern mask, areas adjacent to an irradiation target area may be irradiated with the ion beam, or the edge of the irradiation target area may not be sufficiently irradiated with the ion beam. Japanese Unexamined Patent Application Publication No. 2002-26673 discloses a frequency adjusting method in which the frequency of each piezoelectric element is adjusted by applying an ion beam to a plurality of electrodes formed on a surface of a piezoelectric substrate, and thereby etching the electrodes. This method involves determining a correlation between an ion-beam irradiation time and the amount of frequency change, measuring the frequency of each element on the piezoelectric substrate, determining the amount of frequency adjustment for each element on the basis of a difference between the measured frequency and a target value, determining the ion-beam irradiation time for each element on the basis of the determined amount of frequency adjustment by using the correlation, and applying an ion beam to each element during the determined irradiation time. Japanese Unexamined Patent Application Publication No. 2004-56455 discloses a frequency adjusting apparatus capable of adjusting the frequencies of piezoelectric elements by performing ion beam etching in a vacuum chamber. The frequency adjusting apparatus includes a piezoelectric substrate having a plurality of electrodes formed on its surface; a base plate having an opening for selectively allowing the plurality of electrodes on the piezoelectric substrate to be exposed; an ion source configured to apply an ion beam simultaneously to the plurality of electrodes exposed from the opening in the base plate; a protective plate configured to protect, from being etched, a region around the opening in the base plate to which the ion beam is applied; shutter mechanisms provided as many as the number of the electrodes exposed from the opening in the base plate, and capable of being independently driven; and a masking plate for blocking the ion beam leaking through gaps between the shutter mechanisms. Additionally, as illustrated in FIG. 8, FIG. 9A, and FIG. 9B, there is a conventional frequency adjusting apparatus capable of simultaneously adjusting the frequencies of a plurality of elements arranged perpendicularly to a wafer conveying direction. In this apparatus, a wafer 50, on which a plurality of elements 51 are arranged in a matrix, is conveyed by a conveying unit (not shown) in a wafer conveying direction indicated by an arrow, as shown in FIG. 8. The wafer 50 passes under a pattern mask 52 having a mask hole 53. The mask hole 53 is a slit-like (i.e., elongated) hole allowing a column of elements arranged perpendicularly to the conveying direction of the wafer 50 to be exposed. A plurality of shutters 54 (e.g., six shutters 54 in FIG. 8) are arranged on the pattern mask 52. Each of the shutters 54 is independently actuated with respect to the mask hole 53 in the wafer conveying direction. By selectively covering the slit-like mask hole 53 with the plurality of shutters 54, an irradiation time during which each of the elements 51 is irradiated with an ion beam can be adjusted. In the case of Japanese Unexamined Patent Application Publication No. 2002-26673, a pattern mask having mask holes corresponding to the respective positions of the electrodes is disposed on the piezoelectric substrate. These mask holes are spaced apart from each other, and the ion beam cannot be applied to regions where there are no mask holes. Therefore, this pattern mask is not suitable for use in irradiating small elements closely arranged on the piezoelectric substrate. Additionally, since a rotatable disk-shaped shutter having a double-layered structure is used, it takes time to change the shutter position. Moreover, since the shutter may pass through an opening during rotation, the accuracy of frequency adjustment may be degraded. In the case of the frequency adjusting apparatus disclosed in Japanese Unexamined Patent Application Publication No. 2004-56455, the ion beam is blocked by the masking plate disposed directly above a wafer. However, the shape of holes in this masking plate is not suitable for processing small elements arranged closely on the wafer, as in the case of Japanese Unexamined Patent Application Publication No. 2002-26673 described above. In the case of the apparatus illustrated in FIG. 8, the shutters 54 for covering the slit-like mask hole 53 may be arranged side by side with gaps between adjacent ones, as shown in FIG. 9A, or stacked in a vertical direction, as shown in FIG. 9B. In the case of FIG. 9A, the ion beam enters through the gaps between adjacent shutters 54. This may cause adjustment errors when the elements 51 are closely arranged on the wafer 50. In the case of FIG. 9B, since a masking shape is determined not only by the pattern mask 52 but also by the adjacent shutters 54, a masking position (i.e., distance from the wafer 50) is not constant. Since the ion beam tends to spread out, a change in masking position may cause a change in the amount of spreading of the ion beam. As a result, the amount of etching in adjacent areas or the amount of etching at the edge of a target area may be reduced. Accordingly, it is an object of the present invention to provide a frequency adjusting apparatus capable of dealing with a wafer having many elements closely arranged thereon, and capable of simultaneously and accurately adjusting the frequencies of a plurality of elements on the wafer while keeping an ion-beam masking position (i.e., distance from the wafer) constant. To achieve the object described above, according to preferred embodiments of the present invention, a frequency adjusting apparatus includes a conveying unit configured to convey, in one direction, a wafer on which a plurality of elements are closely arranged; an ion gun for etching, the ion gun being configured to irradiates the wafer with an ion beam while the wafer is being conveyed; a pattern mask having a plurality of mask holes allowing only target areas of the wafer to be exposed, the pattern mask being disposed upstream of the wafer in a direction in which the ion beam travels; and a plurality of shutters each being configured to adjust an irradiation time during which a target area is irradiated with the ion beam, and thereby adjust a frequency in the target area. Each of the mask holes in the pattern mask corresponds to one area of the wafer. The mask holes are alternately displaced in a wafer conveying direction in which the wafer is conveyed, and are arranged in a plurality of columns perpendicular to the wafer conveying direction. The shutters are arranged to correspond to the respective mask holes to individually open and close the corresponding mask holes. Frequency adjustment for areas in one column perpendicular to the wafer conveying direction is performed in multiple steps. According to preferred embodiments of the present invention, a frequency adjusting apparatus includes a conveying unit configured to convey, in one direction, a wafer on which a plurality of elements are closely arranged; a frequency-adjusting-material applying unit configured to apply a frequency adjusting material to the wafer while the wafer is being conveyed; a pattern mask having a plurality of mask holes allowing only target areas of the wafer to be exposed, the pattern mask being disposed upstream of the wafer in a direction in which the frequency adjusting material is applied; and a plurality of shutters each being configured to adjust an application time during which the frequency adjusting material is applied to a target area, and thereby adjust a frequency in the target area. Each of the mask holes in the pattern mask corresponds to one area of the wafer. The mask holes are alternately displaced in a wafer conveying direction in which the wafer is conveyed, and are arranged in a plurality of columns perpendicular to the wafer conveying direction. The shutters are arranged to correspond to the respective mask holes to individually open and close the corresponding mask holes. Frequency adjustment for areas in one column perpendicular to the wafer conveying direction is performed in multiple steps. A description of the frequency adjustment performed using the ion beam in the present invention is provided below. As described above, each of the mask holes in the pattern mask corresponds to one area of the wafer. Since the mask holes are arranged such that processing for areas in one column perpendicular to the wafer conveying direction is performed in multiple steps, the ion beam can be blocked at the same distance from the wafer in both the X and Y directions. Thus, the effect of spreading of the ion beam on the wafer is substantially the same in both the X and Y directions. Therefore, it is possible to minimize reduction in the amount of etching in adjacent areas and the amount of etching at the edge of a target area. Since etching can be made uniformly throughout the entire surface of the wafer, it is possible to realize frequency adjustment for small elements, or areas, closely arranged on the wafer. Although one area (e.g., one mask hole) may correspond to one element, higher processing efficiency can be achieved if one area includes a plurality of elements. An optimum area size can be determined on the basis of the balance between the processing efficiency and the level of accuracy necessary for frequency adjustment. When many elements are formed on one wafer, and since a plurality of neighboring elements are substantially the same in terms of variations in frequency, frequency adjustment may be made by applying the same amount of ion beam thereto. The conveying unit intermittently conveys the wafer at a pitch of one area of the wafer. The mask holes of the pattern mask may preferably be alternately displaced in the wafer conveying direction by a distance of one or more elements and arranged in two columns perpendicular to the wafer conveying direction. Thus, by causing the wafer to pass relative to the pattern mask, all the elements or areas may be uniformly adjusted. Additionally, since the shutters are spaced apart in a direction perpendicular to the wafer conveying direction, the dimensions of each shutter can be set such that one mask hole can be completely closed. This may solve the problem of leakage of the ion beam. It may be more preferable that the mask holes of the pattern mask be alternately displaced in the wafer conveying direction by a distance of one or more pitches and arranged in two columns perpendicular to the wafer conveying direction. This may facilitate the arrangement of areas and data processing. The shutters may be preferably divided into a first shutter group for closing the mask holes in a first column perpendicular to the wafer conveying direction and a second shutter group for closing the mask holes in a second column perpendicular to the wafer conveying direction. First actuators configured to drive the first shutter group and second actuators configured to drive the second shutter group may be preferably arranged opposite each other on both sides of the pattern mask in the wafer conveying direction. Although the first shutter group and the second shutter group may be arranged on the same side of the pattern mask, this arrangement may cause interference between adjacent shutters or actuators. When the first and second actuators are arranged on both sides of the pattern mask, it is possible to properly open and close the mask holes while preventing adjacent shutters and adjacent actuators from interfering with each other. In the embodiments described above, each element is formed in advance such that its frequency is lower than a target value. Subsequently, according to the measured frequency of the element, a target area of the wafer may be etched by ion-beam irradiation so that frequency adjustment can be made. On the other hand, each element may be formed in advance such that its frequency is higher than a target value. Subsequently, according to the measured frequency of the element, a frequency adjusting material may be applied to a target area. Alternatively, after measurement of the frequency of each element, elements whose measured frequencies are lower than a target value may be etched with an ion beam, while a frequency adjusting material may be applied to elements whose measured frequencies are higher than the target value. The level of frequency adjustment made by etching or application of the frequency adjusting material may vary depending on the structure of the element. That is, any level of frequency adjustment may be made, as long as a frequency that is shifted in advance can be brought closer to the target value. Thus, according to the preferred embodiments of the present invention, it is possible to minimize gaps or overlaps between adjacent areas so that the entire surface of the wafer can be irradiated with the ion beam. Even when small elements are closely arranged on the wafer, frequency adjustment can be made uniformly regardless the position (i.e., either at the center or edge) in each area. Additionally, since processing for a column of areas arranged perpendicularly to the wafer conveying direction is performed in multiple steps, frequency adjustment for the entire surface of the wafer can be efficiently made by moving the wafer in one direction. Also, according to the preferred embodiments of the present invention, it is possible to minimize gaps or overlaps between adjacent areas so that the frequency adjusting material can be applied to the entire surface of the wafer. Even when small elements are closely arranged on the wafer, frequency adjustment can be made uniformly regardless the position (i.e., either at the center or edge) in each area. Additionally, since processing for a column of areas arranged perpendicularly to the wafer conveying direction is performed in multiple steps, frequency adjustment for the entire surface of the wafer can be efficiently made by moving the wafer in one direction. Other features, elements, characteristics and advantages of the present invention will become more apparent from the following detailed description of preferred embodiments of the present invention with reference to the attached drawings. Preferred embodiments of the present invention will now be described. First Embodiment FIG. 1 to FIG. 4 illustrate an example of a frequency adjusting apparatus according to an embodiment of the present invention. A frequency adjusting apparatus 1 includes an enclosed processing chamber 2 and a vacuum pump 4 connected to one side of the processing chamber 2, with an opening/closing door 3 interposed between the processing chamber 2 and the vacuum pump 4. By driving the vacuum pump 4, the degree of vacuum in the processing chamber 2 can be maintained at a predetermined level. An ion gun 5 for etching is disposed at the ceiling of the processing chamber 2. As illustrated in FIG. 1, the ion gun 5 is capable of emitting an ion beam IB downward at a substantially constant intensity per unit area within a predetermined region. A movable stage (i.e., conveying unit) 6 is disposed at the bottom of the processing chamber 2. A wafer 10, such as a piezoelectric substrate, is positioned and held on the movable stage 6. As illustrated in FIG. 2 and FIG. 3B, areas 10a, each including a plurality of elements grouped together as a single unit, are closely arranged on the wafer 10 in a matrix at a constant pitch P. The movable stage 6 is capable of intermittently conveying the wafer 10 in the direction of arrow at the pitch P of one area 10a. Before the wafer 10 is conveyed to the frequency adjusting apparatus 1, a frequency in each area 10a of the wafer 10 is shifted in advance in a direction opposite a direction in which the frequency is changed by ion beam irradiation. Then, the frequency in each area 10a is adjusted to be closer to a target value by ion beam irradiation. A layer formed on the wafer 10 and etched by ion beam irradiation may be an electrode layer, a protective layer, or a layer of any material, as long as frequency adjustment can be made. The frequency in each area 10a of the wafer 10 is measured in advance, so that the amount of frequency shift relative to the target value is determined. A shutter base 11 is horizontally disposed at a given position above the movable stage 6. The shutter base 11 has an opening 11a at the center thereof. The size of the opening 11a substantially corresponds to that of a region to be irradiated with the ion beam by the ion gun 5. A pattern mask 15 is positioned and secured in the opening 11a of the shutter base 11. The pattern mask 15 protects a specific region of the wafer 10, except target areas, from being irradiated with the ion beam. It is preferable that the pattern mask 15 be supported at a position as close as possible to the wafer 10. As illustrated in FIG. 3A, the pattern mask 15 is provided with a plurality of mask holes 15a and 15b each corresponding to one area 10a of the wafer 10. The size of each of the mask holes 15a and 15b is determined according to the distance between the pattern mask 15 and the wafer 10 and the spreading angle of the ion beam, and is set to be substantially the same as the size of each area 10a. To perform processing for one column of the wafer 10 in two steps, the mask holes 15a and 15b are alternately displaced in a wafer conveying direction (i.e., X direction) and thus arranged in two columns in a direction (i.e., Y direction) perpendicular to the wafer conveying direction. That is, the mask holes 15a in odd-numbered rows and the mask holes 15b in even-numbered rows are displaced from each other in the X direction (i.e., conveying direction). For example, as illustrated in FIG. 3B, when the areas 10a of the wafer 10 are arranged in nine rows (a) to (i), the mask holes 15a on the downstream side (i.e., left side) in the conveying direction correspond to the respective areas 10a in odd-numbered rows (a), (c), (e), (g), and (i), while the mask holes 15b on the upstream side (i.e., right side) in the conveying direction correspond to the respective areas 10a in even-numbered rows (b), (d), (f), and (h). The two columns of the mask holes 15a and 15b are spaced from each other in the X direction by a distance d corresponding to one column of the areas 10a. The distance d between the two columns of the mask holes 15a and 15b may correspond to one column (i.e., one area pitch=Px, Py) or a plurality of columns of the areas 10a, depending on the distance of one area pitch or the like. As illustrated in FIG. 2, a plurality of actuators 12a and 12b are secured to the shutter base 11, with the opening 11a positioned between the actuators 12a and 12b. As illustrated in FIG. 4, shutters 14a are connected to the respective actuators 12a via respective rods 13, while shutters 14b are connected to the respective actuators 12b via respective rods 13. The shutters 14a and 14b are horizontally actuated individually by the respective actuators 12a and 12b. It is desirable that the actuators 12a and 12b be linear actuators, such as solenoids or voice coil motors, capable of moving between two positions at high speed. The shutters 14a and 14b correspond to the mask holes 15a and 15b, respectively, so as to individually open and close the corresponding mask holes 15a and 15b. That is, the shutters 14a on the left side and the shutter 14b on the right side in FIG. 4 are actuated in opposite directions, so that the shutters 14a open and close the mask holes 15a on the left side and the shutters 14b open and close the mask holes 15b on the right side. It is preferable that the shutters 14a and 14b be moved at a position as close as possible to the upper surface of the pattern mask 15. To completely cover each of the mask holes 15a and 15b, a length L and a width W of each of the shutters 14a and 14b are made greater than their corresponding dimensions Px and Py, respectively, of each of the mask holes 15a and 15b. In the example of FIG. 4, only the upper left actuator 12a is actuated, so that the upper left shutter 14a closes the upper left mask hole 15a. A controller 16, as shown in FIG. 1, controls the movement of the movable stage 6 and each of the actuators 12a and 12b such that frequency adjustment for the areas 10a in each column perpendicular to the wafer conveying direction is performed in multiple steps. For each area 10a of the wafer 10, the amount of frequency shift relative to a target value is stored in the controller 16 in association with an irradiation time during which the area 10a is to be irradiated with the ion beam. After the wafer 10 is conveyed to a position below the pattern mask 15, all the shutters 14a and 14b are opened, so that all the areas 10a corresponding to the mask holes 15a and 15b are irradiated with the ion beam through the mask holes 15a and 15b. Then, the shutters 14a and 14b are closed sequentially in order from the one corresponding to the area 10a for which a desired irradiation time (i.e., irradiation time corresponding to the amount of frequency shift) elapses. After all the shutters 14a and 14b are closed, the wafer 10 is conveyed by a distance of one pitch in the conveying direction. Next, referring to FIGS. 3A and 3B, the order of exposure of the areas 10a from the corresponding mask holes 15a and 15b, i.e., the order of ion beam irradiation, will be described. In FIG. 3B, each of circled numbers 1 to 7 in the respective areas 10a indicates the order of exposure. When the leading end (i.e., left end) of the wafer 10 is brought to a position below the pattern mask 15, four areas 10a at A-b, A-d, A-f, and A-h are firstly exposed from the corresponding mask holes 15b on the right side. Thus, the four areas 10a exposed from the four corresponding mask holes 15b are irradiated with the ion beam, and the irradiation time for each of the four areas 10a is adjusted by the corresponding shutter 14b. After all the shutters 14b are closed, the wafer 10 is conveyed by a distance of one pitch in the conveying direction. Then, four areas 10a at B-b, B-d, B-f, and B-h are secondly exposed from the corresponding mask holes 15b on the right side. At this point, since the wafer 10 has not yet reached the mask holes 15a on the left side, none of the areas 10a is exposed from the mask holes 15a on the left side. The four areas 10a exposed from the four corresponding mask holes 15b on the right side are irradiated with the ion beam, and the irradiation time for each of the four areas 10a is adjusted by the corresponding shutter 14b. After all the shutters 14b are closed, the wafer 10 is conveyed by a distance of one pitch in the conveying direction. Then, thirdly, five areas 10a at A-a, A-c, A-e, A-g, and A-i are exposed from the corresponding mask holes 15a on the left side, while four areas 10a at C-b, C-d, C-f, and C-h are exposed from the corresponding mask holes 15b on the right side. The areas 10a exposed from the corresponding mask holes 15a and 15b are irradiated with the ion beam, and the irradiation time for each of the nine areas 10a is adjusted by the corresponding shutter 14a or 14b. The subsequent operations, which are basically the same as that of the third operation described above, will not be described to avoid redundancy. As described above, the pattern mask 15 is provided with the mask holes 15a and 15b, each corresponding to one area 10a, so that frequency adjustment for the areas 10a in one column perpendicular to the wafer conveying direction can be performed in multiple steps. Accordingly, even when the areas 10a are closely arranged, the entire wafer 10 can be uniformly irradiated with the ion beam. Additionally, the mask holes 15a and 15b can be individually opened and closed by the corresponding shutters 14a and 14b, and the masking height (i.e., distance from the wafer 10) can be kept constant. Therefore, it is less likely that areas adjacent to target areas will be irradiated with the ion beam. At the same time, it is possible to solve the problem where the edge of each target area is not sufficiently irradiated with the ion beam. Second Embodiment FIG. 5 illustrates a frequency adjusting apparatus according to a second embodiment of the present invention. In the present embodiment, one area 10a of the wafer 10 includes an array of four by four elements. The areas 10a are displaced, on a row-by-row basis, by a distance of half a pitch corresponding to two elements in the conveying direction. That is, the areas 10a in the even-numbered rows are displaced in the conveying direction by a distance of half a pitch from the areas 10a in the odd-numbered rows. The amount of displacement d1, however, is not limited to a distance of half a pitch that corresponds to two elements. In the pattern mask 15, a column of the mask holes 15a and a column of the mask holes 15b are spaced from each other in the conveying direction of the wafer 10 by a distance d2 smaller than one pitch P. The distance d2, which is equal to the amount of displacement d1 in the present embodiment, may be set to a value determined by the following equation: d2=n×P+d1(n=1, 2, or the like) The wafer 10 is intermittently conveyed relative to the pattern mask 15 in a direction of arrow at a pitch of one area 10a. Each of circled numbers 1 to 3 in the respective areas 10a indicates the order of exposure. First, two columns of elements in each area marked with a circled number 1 are exposed from the corresponding mask hole 15b on the upstream side and frequency-adjusted. Next, the areas 10a marked with a circled number 2 are exposed from the corresponding mask holes 15b and frequency-adjusted. Then, the areas 10a marked with a circled number 3 are exposed from the corresponding mask holes 15a and 15b and frequency-adjusted. Likewise, the areas 10a corresponding to the mask holes 15a and 15b are sequentially frequency-adjusted. The distance d2 between the columns of the mask holes 15a and 15b may correspond to one column of elements, three columns of elements, or the like. That is, the distance d2 may be determined on an element column by element column basis, depending on the arrangement of the areas 10a. Third Embodiment FIG. 6 illustrates a frequency adjusting apparatus according to a third embodiment of the present invention. In the present embodiment, each of a plurality of pattern masks 20 and 21 is used to process areas only in specific rows. Thus, frequency adjustment for areas in each column perpendicular to the wafer conveying direction is performed in multiple steps. The two pattern masks 20 and 21 are spaced from each other by a predetermined distance D in the wafer conveying direction. It is preferable that the distance D be an integral multiple of the area pitch P. The pattern mask 20 is provided with mask holes 20a1, 20a3, 20a5, 20a7, and 20a9 arranged in a staggered manner. The mask holes 20a1, 20a3, 20a5, 20a7, and 20a9 correspond to areas in respective odd-numbered rows (i.e., the first, third, fifth, seventh, and ninth rows). Specifically, the mask holes 20a1, 20a5, and 20a9 in the first, fifth, and ninth rows, respectively, are formed in the same column. This column is spaced from the other column of the mask holes 20a3 and 20a7 in the third and seventh rows, respectively, in the conveying direction (X direction) by a distance d corresponding to one or several columns. The mask holes 20a1, 20a5, and 20a9 are individually opened and closed by corresponding shutters 22a arranged on the left side of the pattern mask 20. The mask holes 20a3 and 20a7 are individually opened and closed by corresponding shutters 22b arranged on the right side of the pattern mask 20. The other pattern mask 21 is provided with mask holes 21a2, 21a4, 21a6, and 21a8 arranged in a staggered manner. The mask holes 21a2, 21a4, 21a6, and 21a8 correspond to areas in respective even-numbered rows (i.e., the second, fourth, sixth, and eighth rows). Specifically, the mask holes 21a2 and 21a6 in the second and sixth rows, respectively, are formed in the same column. This column is spaced from the other column of the mask holes 21a4 and 21a8 in the fourth and eighth rows, respectively, in the conveying direction (i.e., X direction) by a distance d corresponding to one or several columns. The mask holes 21a2 and 21a6 are opened and closed by corresponding shutters 23a arranged on the left side of the pattern mask 21. The mask holes 21a4 and 21a8 are opened and closed by corresponding shutters 23b arranged on the right side of the pattern mask 21. In the present embodiment, in the wafer 10 that has passed through the pattern mask 21, only areas 10b in the even-numbered rows are frequency-adjusted, as indicated by shading in FIG. 7A. Then, the wafer 10 further passes through the pattern mask 20, so that areas 10c in the odd-numbered rows are frequency-adjusted, as indicated by shading in FIG. 7B. Thus, in the wafer 10 that has passed through the two pattern masks 20 and 21, the frequency adjustment for the areas 10b and 10c in all the rows has been completed. In the case of FIGS. 7A and FIG. 7B, since the pattern mask 21, as shown in FIG. 6, is positioned upstream in the wafer conveying direction, the areas 10b in the even-numbered rows are frequency-adjusted first. As will be understood, if the pattern mask 20 is positioned upstream in the wafer conveying direction, the areas 10c in the odd-numbered rows are frequency-adjusted first. In the present embodiment, areas in one column perpendicular to the conveying direction are frequency-adjusted in four steps. Therefore, even when many very small areas are closely arranged in a wafer, adjacent shutters and actuators can be arranged without interference with each other. As described above, the pattern masks 20 and 21 are spaced from each other by the distance D in the conveying direction. Therefore, it is possible to use actuators each having an axial length much longer than the corresponding area size. It will be understood that the number of pattern masks arranged in the conveying direction is not limited to two, but may be three or more. The present invention is not limited to the embodiments described above, and can be variously modified. Although the above embodiments describe the cases where each area of a wafer and each mask hole of a pattern mask are both substantially square in shape, the present invention is also applicable to the case where they are both substantially rectangular in shape. Although the above embodiments describe the cases where a plurality of elements are included in one area, the present invention is also applicable to the case where one area corresponds to one element. Although the above embodiments describe the cases where a plurality of elements are closely formed on a wafer, the present invention is also applicable to the case where a plurality of elements are closely arranged on a tray. Although the above embodiments describe the cases where an ion beam is applied downward in the vertical direction, the present invention is also applicable to the case where an ion beam is applied upward or horizontally. While preferred embodiments of the invention have been described above, it is to be understood that variations and modifications will be apparent to those skilled in the art without departing from the scope and spirit of the invention. The scope of the invention, therefore, is to be determined solely by the following claims.
051065738	abstract	A steam separator for a boiling water reactor includes a pressure vessel and a chimney spaced radiallly inwardly therefrom to define a downcomer therebetween for recirculating water, the chimney being disposed above a reactor core for channeling upwardly therefrom steam voids and water flow. An annular partition wall is spaced radially between the vessel and he chimney to define an annular collection chamber having an inlet for receiving a portion of the steam voids and water flow from the chimney, a steam outlet for discharging the steam voids from the chamber, and a flow outlet for discharging the water flow from the chamber into the downcomer.
052176789	summary	BACKGROUND OF THE INVENTION The present invention relates to a gang control-rod controlling system for simultaneously operating a plurality of control rods, and more particularly to a gang control-rod controlling system and a reactor operation method, which are effective in exchange of a control rod pattern in boiling water reactors. In boiling water reactors, several hundreds of fuel assemblies are accommodated in the reactor and one control rod for controlling reactor power is installed for every four fuel assemblies. In existing boiling water reactors, control rods are operated one by one. More specifically, first, by actuating a control rod select switch, the radial position of a control rod to be operated is selected. Then, by depressing an "insert" or "withdraw" button, the control is inserted or withdrawn to a predetermined axial position. At the start-up of a reactor, reactor power is increased by withdrawing control rods. At this time, it is required to surely restrict a procedure of operating the control rods within a certain allowable limit so that reactivity worth of each control rod will not be too large. The reason of holding the reactivity worth of each control rod low is as follows. Should there occurs such an accident that any control rod is continuously withdrawn by mistake or that any control rod is dislodged to slip off from the reactor, the amount of radioactive material discharged due to damage of fuel assemblies could be kept within an allowable range from the standpoint of safety evaluation if the reactivity worth applied upon such an event is held low. A rod worth minimizer system (hereinafter referred to as an RWM system or simply as RWM) is known as means for preventing control rods from being withdrawn departing from a predetermined control rod pattern. More specifically, such an RWM system functions to monitor respective positions of control rods and, should an operator attempts to select or withdraw the control rod deviating from a predetermined sequence of control rod operations, to issue an alarm or prevent the attempted operation. However, when reactor power is continuously raised in conformity with the RWM rules, monitoring by the RWM is no longer necessary over a certain power level because the reactivity worth of any control rods becomes small regardless of which control rod is selected. That power level at which the RWM is to be released is usually set to 10 % -35 % of the rated power. In other words, the operation of withdrawing control rods in the range below the power level at which the RWM is to be released must follow the predetermined sequence of control rod operations. Note that JP, A, 49-89094 is known as a prior patent relating to the RWM. The procedure of withdrawing control rods that can hold the reactivity worth of each control rod small is basically to select the control rods to be withdrawn in such order that they are evenly distributed in the radial direction of a core, and also not to successively withdraw those control rods which are adjacent to each other. If two control rods adjacent to each other are withdrawn in succession, the power density at that position would be so extremely increased that the reactivity worth of those control rods becomes too large. Two A type and B type sequences are known as the procedure of withdrawing control rods that can follow the RWM rules. The A type sequence is utilized to configure an A type control rod pattern and the B type sequence is utilized to configure a B type control rod pattern. The term "A type control rod pattern" is here used to mean a pattern of only those control rods arranged in the form of a checker board including a control rod at the core center. The term "B type control rod pattern" is here used to mean a pattern of only those control rods arranged in the form of a checker board in which a control rods at the core center is not included. The A type sequence is set such that the control rods (the B-group control rods) of about half the number of total control rods arranged in the form of a checker board in which a center control rod is not included are divided into four groups, i.e., groups 1 to 4, in such a manner that the control rods of the respective groups are evened in number and arrangement in order to follow the RWM rules so that the reactivity worth of any withdrawn control rod may be held below a certain reference value, and the remaining about half control rods (the A-group control rods) are divided into 18 groups, i.e., groups 5 to 22, each comprising one, four, eight or twelve control rods, in consideration of symmetry. Also, the B type sequence is set such that the control rods (the A-group control rods) of about half the number of total control rods arranged in the form of a checker board including a center control rod are divided into four groups, i.e., groups 1 to 4, in such a manner that the control rods of the respective groups are evened in number and arrangement in order to follow the RWM rules so that the reactivity worth of any withdrawn control rod may be held below a certain reference value, and the remaining about half control rods (the A-group control rods) are divided into 18 groups, i.e., groups 5 to 22, each comprising two, four, eight or twelve control rods, in consideration of symmetry. In the case of configuring the A type control rod pattern by using the A type sequence, to follow the RWM rules, the control rods of the groups 1 to 4 are first operated one by one to be fully withdrawn. As a result, about half the control rods in the entire core are withdrawn in the form of a checker board. This means that whichever control rod is selected at the time of subsequently withdrawing any control rod in the groups 5 to 22, all the control rods adjacent thereto have already been withdrawn and, therefore, the reactivity worth of each control rod is held small. In the above process, the control rods of the same group are always operated to position at the same axial level. Specifically, when some one control rod is withdrawn to a certain axial position, any control rod belonging to other groups shall not be withdrawn until all the remaining control rods of the same group are withdrawn to the same axial position. Also, the B type control rod pattern is configured in a like manner by using the B type sequence. In practical use, only one of the A type sequence and the B type sequence is usually stored in a storage of a central processing unit. Thus, the A type sequence is stored in the storage when the A type control rod pattern is to be configured, and the B type sequence is stored in the storage instead of the A type sequence when the B type control rod pattern is to be configured. While in existing boiling water reactors control rods are withdrawn one by one in conformity with the RWM rules as stated above, gang operation of withdrawing several control rods has been proposed in recent years. Adopting such gang operation for control rods enables cut-down of a time required for the control rod operations and hence a start-up time. During the gang operation, it is also required to follow the RWM rules in the stage of low power. Accordingly, gang control-rod groups in each of which control rods are operated at the same time must be defined in such a manner as able to follow the RWM rules. When the gang operation is adopted to simultaneously withdraw several control rods, it is assumed that those control rods which are simultaneously withdrawn belong to the same control group. In the case of configuring the A type control rod pattern, the aforesaid A type sequence is utilized to select the control rod group to follow the RWM rules as with the existing scheme to operate control rods one by one. In the A type sequence, the number of control rods for each of the groups 1 to 4 is about 1/8 of the total number of control rods in the core. By setting the number of ganged control rods in the groups 1 to 4 so large, the start-up time can be cut down. Specifically, the ganged control rods of the groups 1 to 4 are first fully withdrawn group by group. After that, the ganged control rods of the groups 5 to 22 are withdrawn group by group in view of power distribution across the core since the reactor power is increased. As a result, the A type pattern is configured as a final objective pattern. Also, in the case of configuring the B type control rod pattern, the aforesaid B type sequence is utilized in a like manner. Meanwhile, reactors usually keep on operating for approximately one year, but the control rod pattern is required to be exchanged several times during the continued operation. Let it now be supposed that the control rod pattern is exchanged from the A type to the B type. In this control rod pattern exchange, the control rods used in the B type pattern is first inserted to lower the power level, and the control rods used in the A type pattern is then withdrawn to raise the power level. On this occasion, the steps of inserting and withdrawing the control rods are not carried out at a time, but repeated several times so that the reactor power will not be extremely decreased. In other words, that step of operating the control rods is performed above the power level at which the RWM is to be released. In the existing scheme, since the control rods are operated one by one above the power level at which the RWM is to be released, the aforesaid control rod pattern exchange is carried out by first indicating the radial control rod position to select the control rod to be operated, and then actuating an "insert" or "withdraw" button. However, the following problem arises when the gang operation of control rods is adopted. When inserting the B type pattern control rods, the B type sequence must be selected as a sequence of control rod operations rather than the A type sequence. The reason is that if the A type sequence is selected, even those control rods which must be kept fully withdrawn would be inserted through the gang operation whichever one of the control rod groups 1 to 4 is selected. To the contrary, when withdrawing the A type pattern control rods, the A type sequence must be selected for the same reason. Thus, it is required during exchange of the control rod pattern to repeat several times the steps of inserting the A type pattern control rods and withdrawing the B type pattern control rods, which necessitates one of the B type sequence and the A type sequence to be selected for each of the steps. In the above process, the A type sequence is practically selected by storing the A type sequence in the storage of the central processing unit, and the B type sequence is also selected by storing the B type sequence in the storage. This means that each time the other sequence is selected, the sequence currently stored in the storage requires to be changed, resulting in the very complicated operation. On the other hand, control rods are quite important as means for controlling reactivity of reactors and required to have high reliability in operation. This necessitates that in the gang control-rod operation to operate a plurality of control rods at the same time, only those control rods which are designated as belonging to the same group are surely operated at the same time. Whenever the stored sequence is changed, therefore, it is indispensable to confirm whether the newly stored sequence is correct, or whether any error has occurred. Alternate selection of the A type sequence and the B type sequence entails confirmation of the newly stored sequence whenever it is stored, which makes the operation more complicated and deteriorates the reliability. SUMMARY OF THE INVENTION An object of the present invention is to provide a gang control-rod controlling system and a reactor operation method, which can facilitate the operation of exchanging a control rod pattern and ensure high reliability in the gang control-rod operation to operate a plurality of control rods at the same time. To achieve the above object, in accordance with the present invention, there is provided a gang control-rod controlling system for operating a plurality of control rods at the same time to attain a predetermined control rod pattern, wherein said system comprises (a) first means storing a first sequence of control rod operations for operating a plurality of control rods at the same time to configure a first control rod pattern, a second sequence of control rod operations for operating a plurality of control rods at the same time to configure a second control rod pattern, and a third sequence of control rod operations for operating a plurality of control rods at the same time to exchange a control rod pattern between said first control rod pattern and said second control rod pattern; and (b) second means for selecting one of said first to third sequences of control rod operations stored in said first means, and operating a plurality of control rods at the same time based on the selected sequence of control rod operations. In the gang control-rod controlling system of the present invention thus arranged, the first sequence of control rod operations and the second sequence of control rod operations are utilized to follow the rules of an RWM system when reactor power is low at the start-up of a reactor. In exchanging the control rod pattern, the third sequence of control rod operations is added in view of the fact that the reactor power is above a set level at which the rules of the RWM system are to be released. By selecting the third sequence of control rod operations, further selection of the sequence is no longer required during the exchange of the control rod pattern. Since those sequences are stored in advance, selection of the third sequence of control rod operations only requires an operator to actuate a switch and select it. As a result, the operation of exchanging the control rod pattern is facilitated and high reliability is ensured in the gang control-rod operation. In the above gang control-rod controlling system, preferably, said second means includes means for inhibiting selection of said third sequence of control rod operations when the reactor power is below the set value. With the provision of such means, should the operation of selecting the third sequence of control rod operations is made when the reactor power is below the set level at which the RWM rules are to be released, that selection would not be allowed, making it possible to avoid such a risk that adjacent control rods may be withdrawn in succession below the release power level of the RWM rules, and thus to ensure a high degree of safety. Also preferably, said second means includes rod worth limiting means that functions when the reactor power is below the set value whereby withdrawal of control rods is inhibited when reactivity worth of those control rods exceeds a predetermined range. In this case, more preferably, said second means further includes means actuatable by an operator for selecting the group number of plural control rods to be operated at the same time, and when said reactor power is below the set value, said rod worth limiting means determines whether or not said control rod group number selected by the operator is in match with said selected sequence of control rod operations and, if the decision is no, inhibits output of said selected sequence of control rod operations. By so arranging the second means, when the reactor power is below the set value of the rod worth limiting means in an attempt of selecting the first sequence of control rod operations or the second sequence of control rod operations to start up the reactor, the rod worth limiting means functions in such a manner that even if the operator should select the incorrect control rod group number by mistake, withdrawal of the control rods of that group is inhibited and the reactivity worth of any withdrawn control rod is held within a predetermined range. Preferably, said second means includes first operating means actuatable by the operator for selecting one of said first to third sequences of control rod operations, second operating means actuatable by the operator for selecting the group number of plural control rods to be operated at the same time, third operating means actuatable by the operator for selecting insertion or withdrawal of plural control rods to be operated at the same time, sequence select means for selecting one of said first to third sequences of control rod operations stored in said first means that corresponds to the sequence of control rod operations selected by said first operating means, gang control-rod select means for selecting position data associated with control rods of the group number selected by said second operating means, from the sequence of control rod operations selected by said sequence select means, and control rod operation select means for determining, based on the selection by said third operating means, insertion or withdrawal of the plural control rods associated with the position data selected by said gang control-rod select means. In this case, more preferably, said second means further includes means for, when said reactor power is below the set value, determining whether or not the control rod group number selected by said second operating means is in match with the sequence of control rod operations selected by said sequence select means and, if the decision is no, disabling said selected sequence of control rod operations. In the above gang control-rod controlling system, more preferably, said first sequence of control rod operations includes first sequence data to configure a first type control rod pattern in which control rods inserted at the rated power comprise only those control rods arranged in the form of a checker board including a control rod at the core center, said second sequence of control rod operations includes second sequence data to configure a second type control rod pattern in which control rods inserted at the rated power comprise only those control rods arranged in the form of a checker board in which a control rod at the core center is not included, and said third sequence of control rod operations includes third sequence data in combination of a part of said first sequence data and a part of said second sequence data. In this case, still more preferably, each of said first to third sequences of control rod operations includes data comprising the group numbers of plural control rods to be operated at the same time in correspondence to coordinate values indicative of respective radial positions of the control rods belonging to each group number. Further, said first sequence of control rod operations may contain sequence data in which all control rods are divided into first to m-th groups in order of withdrawing the control rods, the first to fourth groups including those control rods of about half the number of total control rods that are arranged in the form of a checker board and in which a control rod at the core center is not included, and the fifth to m-th groups including the remaining about half control rods; said second sequence of control rod operations may contain sequence data in which all control rods are divided into first to n-th groups in order of withdrawing the control rods, the first to fourth groups including those control rods of about half the number of total control rods that are arranged in the form of a checker board including the control rod at the core center, and the fifth to n-th groups including the remaining about half control rods; and said third sequence of control rod operations may contain sequence data in combination of the fifth to m-th groups in said first sequence of control rod operations and the fifth to n-th groups in said second sequence of control rod operations. To achieve the above object, in accordance with the present invention, there is also provided a reactor operation method for operating a plurality of control rods at the same time to attain a predetermined control rod pattern, wherein said method comprises (a) a first step of previously storing a first sequence of control rod operations for operating a plurality of control rods at the same time to configure a first control rod pattern, a second sequence of control rod operations for operating a plurality of control rods at the same time to configure a second control rod pattern, and a third sequence of control rod operations for operating a plurality of control rods at the same time to exchange a control rod pattern between said first control rod pattern and said second control rod pattern; and (b) a second step of selecting one of said first to third sequences of control rod operations stored in said first step, and operating a plurality of control rods at the same time based on the selected sequence of control rod operations. In the above reactor operation method, preferably, said second step selects one of said second and third sequences of control rod operations when reactor power is below a set value, and selects one of said first, second and third sequences of control rod operations when the reactor power is above the set value. Further preferably, said second step selects the group number of plural control rods to be operated at the same time in response to actuation by an operator, outputs said selected sequence of control rod operations only when said control rod group number selected by the operator is in match with said selected sequence of control rod operations when said reactor power is below the set value, and always outputs said selected sequence of control rod operations when said reactor power is above the set value.
050864432	description	DETAILED DESCRIPTION OF THE INVENTION Briefly, the present invention includes the use of a multiple-layer "wavetrap" deposited over the surface of a layered, synthetic-microstructure soft x-ray mirror optimized for reflectivity at chosen wavelengths for reducing the reflectivity of undesired, longer wavelength incident radiation thereon. In three separate mirror designs employing an alternating molybdenum and silicon layered mirror structure overlaid by two layers of a molybdenum/silicon pair antireflection coating, reflectivities at the wavelengths 133, 171, and 186 .ANG. have been optimized, while that at 304 .ANG. has been minimized. The optimization process involves the choice of materials, the composition of the layer/pairs as well as the number thereof, and the distance therebetween for the mirror, and the simultaneous choice of materials, the composition of the layer/pairs, their number and distance for the "wavetrap." Many details of the present invention are disclosed in the journal article entitled "Metal Multilayer Mirrors For EUV/Ultrasoft X-Ray Wide-Field Telescopes," by Barnham W. Smith, Jeffrey J. Bloch, and Diane Roussel-Dupre, Optical Engineering 29. 592 (1990), the teachings of which are hereby incorporated by reference herein. Reference will now be made in detail to the present preferred embodiment of the invention, an example of which is illustrated in the accompanying drawings. Turning now to FIG. 1 hereof, there is plotted the performance of layered, synthetic-microstructure mirrors showing the effects of employing the "wavetrap" of the present invention versus varying the composition of the layers of the mirror. A.OMEGA. represents a telescope's total area-solid-angle product, and is a measure of a multilayer mirror's performance as it operates within a telescope. To determine A.OMEGA. of a telescope at a given wavelength, the differential contribution of incident rays reflected from the mirror multiplied by the multilayer reflectivity curve for the mirror must be integrated as a function of incidence angle for that wavelength. The goal is to maximize the reflectivity at 171 .ANG., but minimize the reflectivity for the background radiation at 304 .ANG.. Therefore, improved mirror designs are to be found toward the left and top of the plot. FIG. 1a shows the theoretical performance of a molybdenum/silicon multilayer mirror without the "wavetrap" of the present invention as the thickness of the molybdenum layer is varied. .GAMMA. is given by the relationship .GAMMA.=Mo layer thickness/(Mo layer thickness+Si layer thickness). FIG. lb represents the theoretical performance of the 171 .ANG. optimized "wavetrap," while FIG. 1c represents a conservative empirical estimate of how well the "wavetrap" of the present invention will work based on fabricated samples. Optimizations were performed to determine the layer thicknesses yielding the best compromise of high reflectivity for the chosen soft x-ray wavelengths and low response for selected background radiation having longer wavelength. Multilayer reflectivity models were computed with a computer code uses the complex matrix solution method of M. Born and E. Wolf, Principles Of Ootics, Pergammon Press, London (1959), while optical constants employed for molybdenum and silicon were obtained from D. L. Windt, Appl. Opt. 27. 246 (1988), and E. B. Palik, Handbook Of Optical Constants. Academic Press, New York (1985), respectively. Peak reflectivity for the desired angle .theta. is obtained using the Bragg condition for the working wavelength to initially set the spacing of the Mo/Si layers (Bragg condition: 2dsin.theta.=n.lambda., where d is the total thickness of the Mo and the Si layers in each layer pair, .theta. is the angle of reflection from the surface, n is a positive integer and equals unity in this case, and .lambda. is the wavelength). Further fine tuning is necessary because of refraction, absorption, and atom migration in the interface between the surfaces when the layers are set down, as will be set forth below. While maximizing the reflectivity in each mirror's bandpass, it is necessary to minimize sensitivity to background emissions. As stated above, the most serious background for the ALEXIS telescope system in low earth orbit is the geocoronal emission of ionized helium at 304 .ANG.. This radiation is quite intense, perhaps 10.sup.5 times the signal for which measurements are desired in the soft x-ray region from hot, interstellar plasma and other cosmic sources. Therefore, it is necessary to achieve a rejection ratio of at least 10.sup.6 between 304 .ANG. and the peak wavelength for each mirror. A "wavetrap" consisting of two-layer pairs with a different d spacing from that of the multilayered mirror is deposited on top of the other layers. Generally, mirrors are fabricated with a silicon (low-Z) layer farthest from the incident radiation, and a molybdenum (high-Z) layer facing the source of incident radiation, so that the first layer of the "wavetrap" is deposited directly on the high-Z material of the mirror, and is itself a low-Z material. To suppress reflection at 304 .ANG. , the spacing of these extra two-layer pairs are such that standing-wave patterns are set up which destructively interfere with the reflected wave. The 304 .ANG. radiation is then absorbed within the multilayer. It should be mentioned that experiments using one layer pair and three layer pairs were performed with the result that the former did not yield sufficient rejection of the 304 .ANG. radiation, while the latter exhibited excessive absorption of the wavelength for which the mirror reflectivity was optimized. Therefore, two layer pairs were found to be optimal. Since the destructive standing waves of 304 .ANG. radiation in the top two layers interact with the structure of the multilayer mirror below them as a boundary condition, the exact d spacing of the "wavetrap" and the mirror must simultaneously be optimized. That is, the exact d spacing of the "wavetrap" will be different for each type of mirror. Calculations predict the 186 .ANG. design will have a peak reflectivity of 35% for the 186 .ANG. radiation, and a 304 .ANG. reflectivity of less than 10.sup.-5, compared with a peak reflectivity of 40% and a 304 .ANG. reflectivity of 10.sup.-3 without a "wavetrap." Having generally described the present invention, the following example is provided to more particularly set forth the details of apparatus hereof. EXAMPLE Optimized mirrors for three soft x-ray wavelengths have been designed and fabricated as is illustrated in the Table. Therein, the thicknesses of the high-Z material and low-Z material for both the mirrors and their associated "wave traps are provided". Mirrors were constructed having between sixty and one hundred layers and, as stated above, "wavetraps" were found to optimize at two layer pairs. These mirror designs effectively reject the reflections from 304 .ANG. radiation. TABLE ______________________________________ Mirror: "Wavetrap" Wavelength Mo Si Mo Si ______________________________________ 186 .ANG. 31 .ANG. 70 .ANG. 11 .ANG. 47 .ANG. 170 .ANG. 35 .ANG. 58 .ANG. 11 .ANG. 45 .ANG. 130 .ANG. 28 .ANG. 42 .ANG. 10 .ANG. 45 .ANG. ______________________________________ Calculations for the 186 .ANG. wavelength situation suggested that the molybdenum and silicon thicknesses for the mirror and "wavetrap" should be 38 and 74 .ANG., and 10 and 55 .ANG., respectively, as opposed to the values quoted in the Table. When fabricated, although the mirror having these dimensions had a wavelength for peak reflectivity at 186 .ANG., that for the peak rejection efficiency was not at 304 .ANG.. This is thought to be due to migration of the atoms of one layer into another during the deposition process or surface contamination. That is, the layer thicknesses are not precisely defined. Empirical studies and modification of the calculations provided a route to prediction of the optimized values. FIGS. 2a and 2b show the reflectivity for 130 .ANG. radiation and that for 304 .ANG. radiation, respectively, as a function of incidence angle for a typical mirror. As stated, a major problem in the fabrication of multilayer mirrors according to the teachings of the present invention is the layer-to-layer uniformity of the sputtered layers, since only a well-defined layered structure will provide the constructive interference required for maximum reflectivity. The boundary definition can be determined from the number of satellite peaks observed at Cu--K.sub..alpha.. A typical fabricated mirror may have as many as sixteen higher orders visible in a diffractometry measurement. Therefore, attempts to model the Cu--K.sub..alpha. measurements with more than 0.5 .ANG. or .+-.0.5% deviation of the thickness fails to reproduce the observations, indicating that the mirrors are uniform to within this diagnostic's capabilities. However, the empirical fine-tuning required to optimize the "wavetrap" specifications indicate, that the thicknesses may not be exact. Another possibility is that the optical constants derived from the literature are slightly incorrect. Other issues include the two-dimensional uniformity over the surface of the mirror. Nonuniform distances between the substrate and the sputtering system for curved pieces are the source of nonuniformities in layer thickness. Tests on fabricated mirrors show .+-.1% uniformity in the d spacings over the surface over several centimeters diameter piece, and .+-.1.5% over a 15 cm diameter circle. The foregoing description of several preferred embodiments of the invention has been presented for purposes of illustration and description. It is not intended to be exhaustive or to limit the invention to the precise form disclosed, and obviously many modifications and variations are possible in light of the above teaching. For example, although x-ray mirrors for telescopes have been described herein, it would be apparent to one having ordinary skill in the art of x-ray optics that the teachings of the present invention are applicable to focusing mirrors for x-ray lithography procedures using a free-electron laser where the absence of suitable optics currently requires the use of masks having the same dimensions as the circuit dimensions desired on the final integrated circuit chips. The free-electron laser source presents special problems that are solvable using our invention, since a number of harmonics (longer wavelength) of the soft x-ray wavelength to be generated are also present in the laser output. Removal of these harmonics is essential to provide the high resolution required for current lithography processes, since diffraction problems increase as the wavelength increases. Moreover, materials such as tungsten and carbon are known to have good optical properties in the soft x-ray region of the electromagnetic spectrum and are suitable for fabrication of the multilayer mirrors and "wavetraps" of the subject claimed invention. The embodiments were chosen and described to explain the principles of the invention and its practical application to thereby enable others skilled in the art to best utilize the invention in various embodiments and with various modifications as are suited to the particular use contemplated. It is intended that the scope of the invention be defined by the claims appended hereto.
048428078	claims	1. In a nuclear power plant having a reactor vessel including a wall and a head assembly, a core contained within the reactor vessel which includes an array of neutron-producing fuel elements adapted for cyclical replacement upon exhaustion of the fuel, ex-core detector means for measuring a neutron flux, and a primary biological shield substantially surrounding the reactor vessel thereby forming a reactor cavity between the wall and the shield, a system for monitoring neutron exposure to the reactor vessel, comprising: passive dosimetry means for indicating a neutron dosage accumulated over the fuel cycle at a plurality of different preselected locations axially with respect to the core within the cavity, wherein said indicating means is supplemental to the ex-core neutron detecting means; and means for remotely positioning said indicating means accurately and repeatedly at said same plurality of different preselected locations within the cavity, said remote positioning means disposed substantially at a plane where the head assembly joins the wall. at least one dosimeter; and means for housing said at least one dosimeter. means for locating a plurality of predetermined heights relative to the reactor vessel; and means for collecting said indicator means. a plurality of plates, each said plate being mounted within the cavity at a respective predetermined height; and means, attached to each said plate for guiding said transfer means. a pair of brackets, each said bracket including a hole; and a U-shaped tube, each leg of said tube extending upwardly through and attached to a respective one of said holes. a length of beaded chain threaded through said tube; a pair of stop elements, each said stop element being attached to a respective end of said chain thereby limiting the movement of said chain through said tube between a deployed position corresponding to said predetermined heights and a collecting position adapted for retrieval of said sensors sets, wherein said stop elements are further attached to said sensor sets thereby forming a continuous loop; and means for holding said loop in a selected position. a pair of frame tubes; a pair of cross members each of which are attached to said frame tubes thereby forming a substantially rectangular frame assembly; and means for maintaining said frame assembly in a substantially upright position within the cavity. a slit formed in one of said pair of cross members, said slit being adapted to contain a portion of one end of said beaded chain forming said continuous loop; means for suspending said beaded chain contained within said slit; a hole formed in said one of said pair of cross members through which the other end of said continuous loop is adapted to be fed; and a chain support plug to which said other end of the continuous loop is attached, said chain support plug being adapted to engage said hole thereby fixing said continuous loop with said sensor sets at a preselected axial position relative to said core. a frame sleeve coaxially coupled about a respective one of said frame tubes; a joint block coupled to said frame sleeve; a pivot arm tube coupled to said joint block; a pivot arm slide coaxially coupled for sliding engagement about said pivot arm tube; and means for locking said frame sleeve in position with its attached pivot arm tube disposed perpendicular to said frame assembly. a pair of diametrically opposed grooves formed at one end o said frame sleeve; a pair of diametrically opposed pins operable by spring means to extend outward from said frame tube at an upper end thereof, said pins being adapted to fit within said grooves; and means for biasing said frame sleeve upward along said frame tube, thereby engaging said pins within said grooves when said pivot arm tube is disposed perpendicular with respect to said frame assembly. (a) forming a plurality of neutron sensor sets adapted to indicate a neutron dosage accumulated over the fuel cycle; (b) remotely positioning said sensor sets at a plurality of preselected locations within the cavity at a plane where the head assembly joins the wall; (c) exposing said remotely positioned sensor sets by operating the plant through a fuel cycle; (d) remotely retrieving, upon cessation of plant operations following a said fuel cycle, said exposed sensor sets from said plane; (e) conducting neutron activation analysis of said exposed sensor sets; (f) replacing a like plurality of neutron sensor sets at said plurality of preselected locations; and (g) again operating said plant through a fuel cycle and retrieving the exposed sensor sets for analysis. (a) providing a support stand having a generally rectangular frame assembly and pivotable means for holding said frame assembly in a substantially upright position, said holding means including means for locking same in a pivoted position across the cavity; (b) providing a chain to suspend the sensor set; (c) providing a generally U-shaped tube at a location beneath said support stand and the preselected height, the ends of said U-shaped tube extending upward towards said support stand; (d) routing one end of said chain through said U-shaped tube; (e) attaching the other end of said chain to the sensor set; (f) forming a continuous loop with said chain and the sensor set; (g) providing means attached to said continuous loop for indicating when the sensor set is disposed at the preselected height; (h) providing means for holding said continuous loop to said support stand; (i) rotating said continuous loop within said Ushaped tube until such point that said indicating means is proximate to said holding means; and (j) engaging said holding means. 2. The system according to claim 1, wherein said indicating means comprises a plurality of neutron sensor sets. 3. The system according to claim 2, wherein each said sensor set comprises a passive dosimeter. 4. The system according to claim 2, wherein each said sensor set comprises: 5. The system according to claim 4, Wherein said at least one dosimeter is selected from the group of radiometric monitors and solid state track recorders. 6. The system according to claim 2, wherein said indicating means further comprises a plurality of gradient chains connecting said sensor sets, said gradient chains being adapted to react with iron, nickel and cobalt. 7. The system according to claim 6, wherein said gradient chains each comprise a predetermined length of beaded chain formed from an alloy of stainless steel. 8. The system according to claim 1, wherein said remote positioning means comprises: 9. The system according to claim 8, wherein said means for locating comprises means for transferring said indicating means from said preselected locations to a collection location, said transfer means being adapted to minimize interference with a refueling operation undertaken at the end of the fuel cycle. 10. The system according to claim 9, wherein said locating means comprises: 11. The system according to claim 10, wherein said guiding means comprises: 12. The system according to claim 11, wherein said locating means comprises: 13. The system according to claim 12, further comprising a support stand installed within the cavity above said predetermined heights. 14. The system according to claim 13, wherein said support stand comprises: 15. The system according to claim 14, wherein said holding means comprises: 16. The system according to claim 15, wherein said suspending means comprises a spring-loaded plunger having a plunger portion extendable across and above said slit. 17. The system according to claim 14, wherein said maintaining means comprises a pair of pivotable arm assemblies coupled to said frame assembly. 18. The system according to claim 17, where each said pivotable arm assembly comprises: 19. The system according to claim 18, wherein said locking means comprises: 20. The system according to claim 18, further comprising means for biasing said pivot arm slide outward from its respective pivot arm tube. 21. The system according to claim 20, wherein said biasing means comprises a spring installed within said pivot arm slide. 22. The system according to claim 21, further comprising bayonet means for locking said pivot arm slide in a position inwardly along said pivot arm tube, said bayonet means in said position compressing said spring. 23. The system according to claim 18, further comprising a radial take-up bolt threadedly coupled to said joint block opposite said pivot arm tube, said bolt being adjustably positioned inwardly and outwardly from said joint block to maintain said frame assembly in a substantially upright position. 24. The system according to claim 12, further comprising an identification tag attached to said chain, said identification tag including information relating to the plant, the location of said sensor sets within the plant corresponding to a particular azimuth, and the date of dosimetry installation. 25. A method of monitoring neutron exposure to a reactor vessel having a wall and a head assembly in a nuclear power plant having a core contained within said vessel which includes an array of neutron-producing fuel elements adapted for cyclical replacement upon exhaustion of the fuel, and a primary biological shield substantially surrounding said vessel thereby forming a reactor cavity between said vessel and said shield, wherein the method comprises the steps of: 26. The method as described in claim 25, wherein said step of remotely positioning comprises affixing said sensor sets to a chain and positioning said chain in a predetermined position within said cavity. 27. The method as described in claim 26, wherein said step of retrieving comprises moving said chain to an accessible position at said plane and removing said sensor sets from said chain. 28. A method of supporting a radiation sensor set accurately positioned at a preselected height within the cavity of a pressurized water reactor of a nuclear power plant, the reactor having a reactor vessel for containing the core which includes walls and a head assembly, said method comprising the steps of:
	claims	1. A belt for measuring physical quantities of an object, the belt comprising:at least one measurement sensor,a strip having a circumference intended to surround the object,a device for clamping the strip around the object,wherein the belt further comprises a pressing device for pressing the measurement sensor in a first orientation of a first direction directed toward the object,the pressing device comprising at least one casing attached to the strip, at least one intermediate part housed in the casing, and at least one constraining member inserted between the casing and the intermediate part and capable of having the intermediate part assume a first low position in which it presses toward the sensor in the first orientation of the first direction toward the object,the pressing device of the sensor further comprises a lifting member, for holding the intermediate part in a second lifting position above the first low position in a second orientation of the first direction, opposite the first orientation against the constraining member,the lifting member being actuable from the outside of the casing to have the intermediate part pass from the second lifting position to the first low position in which it presses toward the sensor,wherein the lifting member passes through a first guide provided in the intermediate part and abuts against an abutment of the casing in the second lifting position, the lifting member being capable of being removed from the first guide of the intermediate part to have the intermediate part pass from the second lifting position to the first low position in which it presses toward the sensor. 2. The belt according to claim 1, wherein the first guide comprises in the intermediate part a hole for letting through the lifting member in the intermediate part during its passage into the second lifting position. 3. The belt according to claim 1, wherein the lifting member comprises a wire having at least one end section situated outside the casing to allow the lifting member to be removed. 4. The belt according to claim 1, wherein the measurement sensor is a temperature sensor. 5. The belt according to claim 1, comprising a plurality of measurement sensors distributed along the circumference of the strip as a measurement sensor, the plurality of measurement sensors being associated with a plurality of respective pressing devices having a plurality of lifting members. 6. The belt according to claim 5, wherein the lifting members are attached to one another. 7. The belt according to claim 5, wherein the lifting members are formed by the same wire having at least one end section situated outside the casings to allow the lifting members to be removed. 8. The belt according to claim 1, wherein the constraining member comprises a first spring inserted between the casing and the intermediate part. 9. The belt according to claim 1, comprising at least one heat-insulating layer between the intermediate part and the measurement sensor. 10. The belt according to claim 1, wherein the device for clamping the strip around the object comprises:at least one first hooking part attached in proximity to a first end of the strip and at least one second hooking part attached in proximity to a second end of the strip,a first module for connection to the hooking parts, capable of being mounted removably on them,the first module comprising a first spindle for driving the first hooking part in a first joining direction coming closer to the second hooking part and a second spindle for driving the second hooking part in a second joining direction coming closer to the first hooking part, at least one second guide on which the first and second spindles are slidably mounted respectively in the first and second joining directions, and at least one second bias spring mounted between at least one of the spindles and the second guide to cause the spindles to come closer one to another in the first and/or second joining direction,a second approximation module for bringing the spindles closer in the first and second directions, allowing the immobilization of the spindles in a clamping position of the belt around the object. 11. The belt according to claim 10, wherein the second approximation module comprises a gripper for gripping the spindles. 12. The belt according to claim 10, wherein the second approximation module comprises at least one first jaw for gripping the first spindle and at least one second jaw for gripping the second spindle, the first jaw being integral with at least one first arm, the second jaw being integral with at least one second arm, the first arm being hinged with respect to the second arm by a main axis of rotation situated at a distance from the jaws, the second approximation module further comprising at least one screw cooperating with the arms to cause the jaws to come closer one to another by rotation around the main axis. 13. The belt according to claim 12, wherein the approximation module is of the parallelogram or pantograph type between the screw and the jaws. 14. The belt according to claim 12, wherein the second approximation module comprises at least one first connecting rod having a first hinge axis with respect to the first arm between the main axis and the first jaw, at least one second connecting rod having a second hinge axis with respect to the second arm between the main axis and the second jaw, the connecting rods being mutually hinged by a third axis situated at a distance from the first and second axes, the screw cooperating with a first support mounted on the main axis and with a second support mounted on the third axis to allow the jaws to come closer one to another by moving the first and second supports away one from another. 15. A belt for measuring physical quantities of an object, the belt comprising:at least one measurement sensor,a strip having a circumference intended to surround the object,a device for clamping the strip around the object,wherein the belt further comprises a pressing device for pressing the measurement sensor in a first orientation of a first direction directed toward the object,the pressing device comprising at least one casing attached to the strip, at least one intermediate part housed in the casing, and at least one constraining member inserted between the casing and the intermediate part and capable of having the intermediate part assume a first low position in which it presses toward the sensor in the first orientation of the first direction toward the object,the pressing device of the sensor further comprises a lifting member, for holding the intermediate part in a second lifting position above the first low position in a second orientation of the first direction, opposite the first orientation against the constraining member,the lifting member being actuable from the outside of the casing to have the intermediate part pass from the second lifting position to the first low position in which it presses toward the sensor,wherein the lifting members are attached to one another. 16. A belt for measuring physical quantities of an object, the belt comprising:at least one measurement sensor,a strip having a circumference intended to surround the object,a device for clamping the strip around the object,wherein the belt further comprises a pressing device for pressing the measurement sensor in a first orientation of a first direction directed toward the object,the pressing device comprising at least one casing attached to the strip, at least one intermediate part housed in the casing, and at least one constraining member inserted between the casing and the intermediate part and capable of having the intermediate part assume a first low position in which it presses toward the sensor in the first orientation of the first direction toward the object,the pressing device of the sensor further comprises a lifting member, for holding the intermediate part in a second lifting position above the first low position in a second orientation of the first direction, opposite the first orientation against the constraining member,the lifting member being actuable from the outside of the casing to have the intermediate part pass from the second lifting position to the first low position in which it presses toward the sensor, andwherein the device for clamping the strip around the object comprises:at least one first hooking part attached in proximity to a first end of the strip and at least one second hooking part attached in proximity to a second end of the strip,a first module for connection to the hooking parts, capable of being mounted removably on them,the first module comprising a first spindle for driving the first hooking part in a first joining direction coming closer to the second hooking part and a second spindle for driving the second hooking part in a second joining direction coming closer to the first hooking part, at least one second guide on which the first and second spindles are slidably mounted respectively in the first and second joining directions, and at least one second bias spring mounted between at least one of the spindles and the second guide to cause the spindles to come closer one to another in the first and/or second joining direction,a second approximation module for bringing the spindles closer in the first and second directions, allowing the immobilization of the spindles in a clamping position of the belt around the object. 17. The belt according to claim 16, wherein the second approximation module comprises a gripper for gripping the spindles. 18. The belt according to claim 16, wherein the second approximation module comprises at least one first jaw for gripping the first spindle and at least one second jaw for gripping the second spindle, the first jaw being integral with at least one first arm, the second jaw being integral with at least one second arm, the first arm being hinged with respect to the second arm by a main axis of rotation situated at a distance from the jaws, the second approximation module further comprising at least one screw cooperating with the arms to cause the jaws to come closer one to another by rotation around the main axis. 19. The belt according to claim 18, wherein the approximation module is of the parallelogram or pantograph type between the screw and the jaws. 20. The belt according to claim 18, wherein the second approximation module comprises at least one first connecting rod having a first hinge axis with respect to the first arm between the main axis and the first jaw, at least one second connecting rod having a second hinge axis with respect to the second arm between the main axis and the second jaw, the connecting rods being mutually hinged by a third axis situated at a distance from the first and second axes, the screw cooperating with a first support mounted on the main axis and with a second support mounted on the third axis to allow the jaws to come closer one to another by moving the first and second supports away one from another.
046613110	summary	BACKGROUND OF THE INVENTION The invention concerns a nuclear power plant arranged in an underground cavity comprising a small high-temperature pebble bed reactor; a reactor core; spherical operational elements which pass through the core more than once; a steel reactor pressure vessel in which the small high-temperature reactor and a heat exchange apparatus are installed; and a loading installation for the addition and removal of fuel elements. A nuclear powe plant of this type is described in German patent application No. P 33 35 451.0. In this nuclear power plant, all of the components of the primary loop, together with the control and shutdown installations are disposed within a steel reactor pressure vessel in such a manner that they may be installed and removed from above. This renders an economical underground construction possible. Under the small reactor, at least one discharge tube is provided for the removal of the spherical fuel elements. Another nuclear power plant with a small high-temperature reactor suitable for installation in an underground cavity is described in German patent application No. P 33 45 113.3. Here again, the spherical fuel elements introduced on top are drawn off by means of a discharge device at the bottom of the pile. Loading installations for nuclear reactors of medium capacity with spherical fuel elements which are built by the principle of modular construction are known. These installations comprise movable functional parts for the addition, removal, distribution and extraction of fuel elements by means of drives. The functional parts are provided with bores for the passage of fuel elements and are set into a block, or plate equipped with connecting bores for the fuel elements. Feed installations of this type are described in German Gebrauchsmuster No. 6 753 677, German Auslegeschrift No. 15 89 532 and German Offenlegungsschrift No. 23 57 426. The manner in which the introduction of the fuel elements into the reactor core is effected, is not disclosed in the references cited. The state of the art further includes German Pat. No. 1 281 046, which again concerns the feed installation of a so-called pebble bed reactor of intermediate capacity. The discharge and sorting installation for the spherical fuel elements is located under the reactor core. It contains a measuring device in which the fuel elements are examined with regard to their state of depletion. Depending on the results of the measurement, the fuel elements are either returned into the reactor core or eliminated from the circulation of fuel elements. The addition of fresh fuel elements is effected from above through a feed tube. The installation and removal of primary loop components is thus rendered difficult. SUMMARY OF THE INVENTION It is an object of the present invention to design the feed installation for the small high-temperature pebble bed reactor in a nuclear power plant of the aforedescribed structural type, so that access to the components of the primary loop from above is not impaired, and so that the loading installation occupies as little space as possible. According to the present invention, this and other advantageous objects are attained by providing a divided loading space under the reactor pressure vessel. A discharge installation comprising removal, depletion and sorting devices is located in the loading space, together with a conveyor installation. An addition device for fresh fuel elements and a collector vessel for used fuel elements are arranged outside the underground cavity, and are connected with the lower part of the loading space by means of vertical conduits, mounted outside the reactor pressure vessel. The conveyor installation comprises an inward transfer block for transferring fresh fuel elements from the addition device as well as partially depleted fuel elements sorted out in the depletion measuring installation. The conveyor installation is connected with a distribution installation for the reactor core by a vertical ascending line located above the reactor core, outside and to the side of the reactor pressure vessel. The depletion measuring installation is connected with an outward transfer block for depleted fuel elements by means of another line which is in turn connected with the vertical conduit of the collector vessel. In the nuclear power plant according to the invention, fresh fuel elements are therefore not added directly into the reactor core from above, but are conveyed outside the reactor pressure vessel into the lower part of the loading installation and transported to the distribution device in the same line with the partially depleted fuel elements. From there they are fed laterally into the reactor pressure vessel. Unimpaired access to the primary loop components from above is thus provided. According to a further embodiment of the invention, a column of fuel elements is always present in the conduit in order to prevent damage by impact to the fresh fuel elements in the course of their transport through the vertical conduit. The great fall height thus cannot have a harmful effect on the fuel elements, even in the absence of braking means. The aforedescribed objects of the invention may also be attained with a configuration different from the one described above. One alternative configuration according to the present invention is characterized in that a divided loading space is provided under the reactor pressure vessel; an addition device for fresh fuel elements and a collector vessel for used fuel elements are disposed in the loading space, together with a discharge installation comprising removal, depletion measuring and sorting devices and a conveyor device; the loading space is accessible through a horizontal channel and a vertical shaft; the conveyor installation comprises an inward transfer block for the introduction of fresh fuel elements from the addition device, together with partially depleted fuel elements sorted out in the depletion measuring device; the conveyor installation is connected with a distribution device for the reactor core by means of a vertical ascending line; the distribution device is provided above the reactor core, outside and to the side of the reactor pressure vessel; the depletion measuring device is connected by another line with an outward transfer block for depleted fuel elements, which, in turn, is connected with the collector vessel by a further line. The operating elements of a small high-temperature pebble bed reactor with a capacity of 100 to 200 MWe comprises one-half fuel elements and one-half pure graphite elements, or so-called blind elements. The above is true if one disregards the initial loading phase in which absorber elements are also added. As the consumption of the graphite elements cannot be measured with the depletion measuring installation, it is proposed to completely replace the graphite elements with fresh graphite elements in the course of the operational recirculation of the core load after an operating period of (n-1) years in order to avoid greater expense. The symbol n is defined as the service life of the graphite elements in years. Under the conditions prevailing in a small reactor n amounts to about 10 years. With this method, even though the quantity of fresh graphite elements is twice as high as that actually consumed, there is still a cost saving.
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