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	abstract	In accordance with at least some embodiments of the present disclosure, a process to estimate scattered radiation contained in x-ray projections for computed tomography (CT) reconstruction is provided. The process may construct an object model based on a plurality of projection images generated by CT scanning of an object using an x-ray radiation source and a detector panel. The process may construct a virtual radiation source based on the x-ray radiation source, and a virtual detector panel based on the detector panel. The process may perform a simulated CT scanning of the object model by simulating macroscopic behavior of particles being emitted from the virtual radiation source, passing through the object model, and being detected by the virtual detector panel. And the process may generate a simulated scatter image based on a first subset of particles scattered during the simulated CT scanning of the object model.
	abstract	An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.
	description	This is the U.S. National Stage of International Application No. PCT/US2005/027558, filed Aug. 2, 2005, which was published in English under PCT Article 21(2), which claims priority from U.S. application Ser. No. 10/911,407, filed Aug. 2, 2004, which is incorporated herein by reference. Disclosed are methods and apparatus for producing radioisotopes, such as molybdenum-99, as well as novel compositions produced according to the methods. Radioisotopes have significant utility for a wide range of medical applications. For example, technetium-99m is a primary radioisotope employed in diagnostic nuclear medicine. The importance of technetium-99m to clinical procedures is highlighted by the fact that more than 13 million diagnostic procedures using technetium-99m are performed each year in the United States alone. Technetium-99m has a combination of desirable physical properties, including its gamma decay mode and energy, that are ideally suited for single photon emission computed tomography. Moreover, technetium's chemical reactivity and versatility allow it to be conveniently complexed to carrier or targeting agents, such as antibodies, peptides, and other molecules, which allows particular tissues to be selectively imaged or scanned. Because technetium-99m has a short half-life (t1/2=6 h), this radioisotope, when used in clinical practice, is typically produced from its longer-lived parent nuclide, molybdenum-99 (t1/2=66 h) using a chromatographic column generator. For an example of a molybdenum-99/technetium-99m generator, see U.S. Pat. No. 5,774,782 to Mirzadeh et al. Removing the technetium-99m daughter nuclide from the generator (i.e., separating it from molybdenum) is typically performed by “milking” the generator a few times daily by pulling normal saline through the column to elute the soluble technetium-99m for complexation in “kits” for subsequent patient injection. Commercial quantities of molybdenum-99 have been produced in nuclear reactors over the years through the uranium fission process (for example, see U.S. Pat. No. 3,799,883 to Arino et al.) utilizing highly enriched uranium-235 that requires extensive security and non-proliferation safeguards. Unfortunately, the fission process, whether it is based on low or high enriched uranium, yields only a small amount of molybdenum-99 with a large array of undesirable fission products that present significant infrastructure, health and security, liability, handling, storage, and waste issues and associated costs. Further, this mode of production requires dedicated and reliable nuclear reactors, support facilities and operation thereof to maintain a continuous supply. The United States currently depends solely on a limited number of foreign suppliers of molybdenum-99. Disclosed herein are radioisotope compositions, including molybdenum-99 compositions, as well as novel methods, apparatus and systems for producing such compositions. Some of the disclosed methods for producing molybdenum-99 compositions include providing a zirconium-96-containing starting material, such as a starting material enriched in zirconium-96. The zirconium-96-containing starting material can be manufactured or otherwise formed into a target. The target can be irradiated with charged particles to transmute material in the target into molybdenum-99. The molybdenum-99 then can be separated from the target material. Alternatively, the zirconium-96-containing starting material can be irradiated without being formed into a target prior to irradiation. Certain embodiments of the disclosed methods include removing the irradiated material from the target, for example by ionization, ablation, spallation, sputtering, and/or mechanical removal, such as milling. These and other embodiments also can include partially or completely dissolving the irradiated material to produce an irradiated target solution. Harvesting the irradiated material from the target can be performed during irradiation, following irradiation, or both. Separation of the desired species, such as the desired molybdenum-containing species, from other species can be accomplished by any suitable process, such as chemical separation, mass difference separation, plasma separation, (e.g., diffusion, centrifugation, and/or mass spectrometry) or combinations thereof. In a particular embodiment, harvesting molybdenum-99 includes contacting an irradiated target or material from an irradiated target with a solvent, such as aqua regia, to produce an irradiated target solution containing molybdenum-99 and zirconium. The molybdenum-99 then can be separated from other species in solution, such as zirconium species and incidental impurities or transmutation products. Purification can be accomplished by ion-exchange chromatography. For example, molybdenum-99 can be separated from zirconium by anion-exchange chromatography. Some embodiments of the disclosed methods for making and purifying molybdenum-99 include contacting the irradiated target or material from the irradiated target with a fluorinating agent, such as NF3 and/or HF. Fluorination of such materials can produce MoF5, MoF6 or both, as well as zirconium fluoride species. The molybdenum fluoride products MoF5 and MoF6 are relatively volatile at about ambient temperatures, whereas zirconium fluoride compounds produced by the process are non-volatile at about ambient temperatures. Thus, the desired molybdenum-99 material can be isolated by evaporation and condensation. Upon separation from the zirconium species, molybdenum fluorides optionally can be hydrolyzed to produce molybdate, which is typically the molybdenum species used to produce technetium-99m. Also disclosed are targets for producing the disclosed radiopharmaceutical compositions. In certain embodiments, the targets are enriched in zirconium-96. The targets may take any suitable form to meet operational and production requirements. Certain target embodiments include target shapes and dimensions selected so that product yield and/or purification efficiency are optimized. In certain embodiments, the target comprises a disk, a ribbon, a wire or combinations of these forms. Because of the value associated with the target material, some embodiments are designed to enable the product to be harvested and the target to be subsequently re-irradiated for further product formation and harvesting. One embodiment of a system for producing molybdenum-99 includes an alpha particle source for producing an alpha particle beam and a target comprising zirconium-96 arranged such that at least a portion of the alpha particle beam intersects the target. Any suitable charged-particle beam source can be used, including, without limitation, an alpha particle source comprising a table-top generator, a cyclotron or a linear accelerator. In certain embodiments of the system, a bath is provided for harvesting irradiated material from the target by chemical means. Mass difference means also can be used to harvest irradiated material from the target. Both of these harvesting methods, and several other harvesting methods, are suitable for continuous or near-continuous processing during irradiation, following irradiation, or both. For increased efficiency, the disclosed systems can include means for conducting the irradiation processes in a continuous mode or near-continuous mode by continuously exposing non-irradiated portions of the target (or starting material) to the charged-particle beam, which may be continuous or pulsed with a duty factor. In systems where the target is in a form of a disk or other rotatable shape, the process may be facilitated by rotating the target relative to the beam so that fresh material is thereby exposed. In some embodiments, the process includes harvesting irradiated target material at one location on the target while another portion of the target is being irradiated. Also disclosed are compositions comprising molybdenum-99. For example, certain embodiments of the compositions can include 80% or greater abundances of molybdenum-99, and in some embodiments, the compositions include at least about 90% molybdenum-99. Such compositions also can include a carrier, such as a carrier gas. The carrier gas may comprise, e.g., helium, argon, combinations thereof and other suitable carriers. Embodiments of the compositions disclosed herein include radiopharmaceutical compositions containing one or more of zirconium-96, molybdenum-99 and technetium-99m. In certain embodiments, the compositions include molybdenum-99 produced by the process described above. The molybdenum-99 produced as described herein can be used to prepare a technetium-99m generator. Such generators are suitable for clinical use, for example, for use at hospitals to produce the disclosed radiopharmaceutical solutions. The preparation of the technetium-99m generator typically involves loading the purified molybdenum-99 onto an adsorbent column. After an appreciable amount of molybdenum-99 has decayed into technetium-99m, the technetium can be eluted from the column, thereby separating the molybdenum species from the technetium species. Certain embodiments of the molybdenum and technetium compositions disclosed herein are substantially free of impurities, such as radioactive impurities. Some of the disclosed compositions include less than about 5×10−2% percent radioactive impurities (by weight). Specifically, particular compositions are substantially free of actinides. Similarly, certain compositions are substantially free of radioactive isotopes of strontium, ruthenium, tellurium and iodine. One embodiment consists essentially of zirconium and molybdenum-99. In addition to the desired radioisotope, some embodiments of the disclosed compositions include at least one carrier. For example, radiopharmaceutical compositions and many of their precursor compositions can include a pharmaceutically compatible carrier. Disclosed herein are methods, apparatus and systems for the production of radioisotopes, such as molybdenum-99. Also provided are novel medical radioisotope compositions and their precursors, including compositions comprising molybdenum-99, technetium-99m, or both. It should be understood that when a particular isotope, such as molybdenum-99 or technetium-99m is referred to herein, compounds containing the particular isotope also are intended. With respect to molybdenum-99 and technetium-99m, such compounds can include, without limitation, molybdate (MoO42−) salts, pertechnetate (TcO41−) salts, chlorocomplexes of both and other chemical species. Unless otherwise indicated, all numbers expressing quantities of ingredients, properties, thicknesses, power levels, and so forth used in the specification and claims are to be understood as being modified by the term “about” whether explicitly stated or not. Accordingly, unless indicated clearly to the contrary, the numerical parameters set forth are approximations. The disclosed methods for making radioisotopes may include providing a starting material. For example, the disclosed methods for making molybdenum-99 and technetium-99m may include providing a zirconium-96-containing starting material. Some embodiments of disclosed methods for preparing molybdenum-99 are illustrated in FIG. 1. With reference to FIG. 1, these embodiments begin with a target containing zirconium-96. Irradiation of the target with alpha particles yields, via an alpha particle capture/neutron emission process, an irradiated target containing molybdenum-99. The irradiated target can be processed to harvest the irradiated material and purify the molybdenum-99. Alternatively, the starting material containing zirconium-96 can be irradiated without forming a target. As illustrated in FIG. 1, the starting material can be irradiated directly to produce molybdenum-99. The irradiated starting material also can be processed to purify molybdenum-99 from the remaining components of the starting material. After molybdenum-99 is harvested from the target material or the starting material, it typically is loaded onto a technetium generator from which technetium-99m can be directly extracted by the end user. Typically, the molybdenum-99 product is purified to separate it from zirconium isotopes present in the irradiated target or irradiated starting material prior to loading onto the technetium generator. As is known in the art, technetium-99m can be separated as pertechnetate (99mTcO41−) from its parent hydrated molybdenum trioxide (MoO3) or molybdate ion (99MoO42−) via column chromatography. Examples of technetium generators for purifying radiopharmaceutical quality technetium-99m produced from molybdenum-99 decay are described in U.S. Pat. No. 5,774,782 and in U.S. Patent Pub. No. 2003/0219366, both of which are incorporated herein by reference. Further variations of the processes, as well as other specific embodiments are set forth below. In embodiments of the disclosed processes, the starting materials can be present in various compounds and compositions. For example, the starting materials can be in their natural abundance or enriched. Zirconium has a natural abundance of about 2.8% zirconium-96. Zirconium-containing materials can be used in embodiments of the disclosed methods without enrichment, as shown in Example 4, below. Zirconium-containing materials also can be enriched to higher abundances of zirconium-96. Enrichment, in particular embodiments, improves the yield of molybdenum-99 and, in certain embodiments, reduces the concentrations of undesirable impurities, such as molybdenum-93, in the product. In some embodiments, the concentration of zirconium-96 can range from greater than about 2.8% to greater than about 95%, such as from about 10% to greater than about 95%. Typically, when a target enriched in zirconium-96 is used, the zirconium includes at least about 10% of the zirconium-96 isotope, such as at least about 50% and, in some embodiments, at least about 90%. Methods for preparing compounds enriched in zirconium-96 are known to those of ordinary skill in the art. For example, plasma separation is one technique that can be used to separate zirconium-96 from other zirconium isotopes. Plasma separation methods and apparatus are well known to those of ordinary skill in the art. For examples, see, Rosenthal et al., Localized Density Clumps Generated in a Magnetized Nonneutral Plasma. Phys. Lett. A 1992, 170, 443-447; Bauer et al., Experimental-Observation of Superstrong Electron-Plasma Waves and Wave Breaking. Phys. Rev. Lett. 1992, 68, 3706-3709; and U.S. Pat. No. 5,981,955 to Wong and Rosenthal. Zirconium-96 also can be obtained from spent nuclear fuel. In addition, zirconium compounds enriched in zirconium-96 are commercially available. For example, zirconium oxide enriched in zirconium-96 is commercially available at various enrichment levels. Zirconium oxide having 58.5% zirconium-96 is commercially available, for example, from STB Isotope, Hamburg, Germany, and zirconium oxide having over about 95% zirconium-96 is available, for example, from International Isotopes Clearing House Inc., Leawood, Kans., and from Chemotrade, Düsseldorf, Germany. Any zirconium-96-containing starting material can be irradiated to produce molybdenum-99. Suitable starting materials include, without limitation, zirconium metal, zirconium nitride and zirconium oxide. Zirconium oxide is useful due to its ready availability, low cost and chemical stability. Zirconium metal also is readily available and, due to its ductility and malleability, is useful for fabricating into a desired target form. The starting material optionally can include other elements or compounds, such as other metals in addition to zirconium, to facilitate target construction. The starting material can be fabricated into various target configurations to enhance the production and recovery of the desired species, such as molybdenum-99. These targets can be used in conjunction with an alpha-particle source. FIG. 2 illustrates one embodiment of a system 10 for producing radioisotopes, including a source 20, a beam 30 and a target 40. The target can be in gaseous, liquid or solid form. Typically, the target is designed or adapted to dissipate excess heat and/or charge that builds up during irradiation. For example, the target can have a relatively large surface-area-to-volume ratio to favor heat dissipation and/or be thermally coupled to a heat-rejection medium, such as a gas (e.g., helium) or a liquid (e.g., water). Phase change reactions in the heat-rejection medium are particularly useful for absorbing heat generated by the irradiation process. In some embodiments, the heat-rejection medium is in the form of a mesh or screen comprising a material with high thermal conductivity. In these embodiments, the starting material can be, for example, placed in the interstices of the mesh or screen. Those skilled in the accelerator art know of additional methods for dissipating heat from the target, e.g. by design of the target. See, for example, Larsen et al., Evaluation of an Internal Cyclotron Target for the Production of 211At via the 209Bi (α,2n) 211At Reaction. Appl. Radiat. Isot. 1996, 47, 135-143. To dissipate charge, the target can be, e.g., electrically coupled to ground. The beam energy and target thickness may be selected such that the efficiency or economics of the overall process is optimal. For example, the path length of alpha particles within zirconium-96-containing materials is dependent upon the beam energy. Specifically, the path length of a 15 MeV beam of alpha particles within such materials is about 200 μm. In certain embodiments, the target is designed to have a thickness of about the alpha particle path length or less, such that the target is irradiated to its full thickness. In some such embodiments purification of the desired product is more efficient because less starting material remains in the target. In some embodiments, however, only the irradiated material is removed from the target for purification, which also simplifies purification of the desired product. In certain embodiments, the target has a thickness substantially parallel relative to the beam vector or axis of less than about 1 mm and more typically has a thickness of from about 0.05 mm to about 0.5 mm. In particular embodiments, the target thickness is from about 0.2 to about 0.4 mm or from about 0.1 to about 0.2 mm. Thicker targets also can be used. Thicker targets are especially well-suited for dissipating heat generated by irradiation and for facilitating harvest of the desired product by mechanical processes. In some embodiments, the target has a thickness greater than about 1 mm, such as from about 1 mm to about 50 mm, from about 20 mm to about 40 mm, or from about 10 mm to about 20 mm. The desired thickness depends in part upon the energy level of the particle beam. Similarly, in some embodiments, the target has a width substantially perpendicular to the beam axis of less than about the beam diameter. Such dimensions are designed to result in a greater percentage of the target material being irradiated. As discussed above, this may simplify purification by reducing the amount of starting material that needs to be separated following irradiation. In one embodiment of a system for producing molybdenum-99, the alpha particle beam diameter is less than about 1 mm. The alpha particle beam diameter also can be greater than about 1 mm. One embodiment of a target well-suited for dissipating heat is depicted in FIG. 3A. The illustrated target assembly 50 includes a target formed into the shape of or onto a disk 70, which in turn can be mounted on a spindle 60. With reference to FIGS. 3A and 3B, the disk 70 includes an irradiation surface 80 arranged in an alpha beam path 90. Rotation of the disk 70 about the spindle 60 may place fresh portions of the irradiation surface 80 in the beam path 90. The spindle 60 optionally includes a ground connection (not shown) thus providing electrical coupling of the disk 70 to ground to dissipate charge accumulated during irradiation. Typically, the disk 70 has a diameter of from about 1 cm to about 30 cm, and in one embodiment, the disk 70 has a diameter of about 30 cm. Of course, the dimensions of the disk 70 are based on the other various parameters of the apparatus. In another embodiment, the disclosed target can take the shape of or be formed into a ribbon, the ribbon, for example, being similar in shape to the shape of a piece of photographic film. The ribbon can be passed through an alpha particle beam at a rate suitable for producing useful amounts of the product, such as molybdenum-99. In this embodiment, the ribbon can, for example, be wound from one spool to another. As with the disk-shaped target discussed above, the target ribbon is well-suited to a continuous process wherein irradiated target material is harvested from an irradiated portion of the target ribbon while a second portion of the target ribbon is being irradiated. The ribbon can have a thickness similar to that discussed above in relation to the disk-shaped target, for example, 200 μm. The width of the ribbon may be comparable to the width of the alpha beam. In certain embodiments, the target includes a substrate of a first material and is coated or in some manner combined with a layer of the starting material, such as a layer of zirconium-containing material. The substrate may comprise a flexible or rigid material depending upon the form the target is to take. The substrate material can be selected to have sufficient thermal conductivity to withstand the heat produced during alpha particle irradiation and can be selected to readily dissipate heat. The substrate also can be selected so that it does not produce undesirable transmutation products during irradiation. The layer of zirconium-containing material can have a thickness similar to that discussed above in relation to the disk-shaped target. In some disclosed embodiments, the substrate is coated with at least about 200 μm of a starting material, such as from about 200 μm to about 1 cm. The coating can have a thickness, for example, sufficient to substantially prevent alpha particles from reaching the substrate. The particle beam energy and other parameters can be set accordingly. In certain embodiments, such as where the target is a ribbon, the substrate can be selected such that the substrate and coating are sufficiently flexible to be wound about a spool. The substrate can be an inorganic material or materials, such as a metal, and/or can be an organic material or materials, particularly a synthetic material, such as a synthetic polymer. Embodiments containing a substrate and a coating optionally can include an interlayer positioned between the substrate and the coating. In some embodiments, the interlayer is selected to improve adhesion between the substrate and the coating. Thus, such interlayers may be employed, for example, when the substrate and the coating do not adhere with sufficient affinity. In such cases, interlayer materials can be selected by a person of ordinary skill in the art such that the interlayer adheres to the substrate and the coating, thus effectively bonding the substrate and the coating. Combinations of coatings, substrates and interlayer materials can be selected for compatibility and appropriate physical properties by a person of ordinary skill in the art. A variety of alpha particle sources are compatible with the disclosed methods. For example, a zirconium target can be irradiated with alpha particles using any alpha particle source to produce molybdenum-99. Alpha particle sources include cyclotrons, linear accelerators and table-top generators. Examples of suitable cyclotrons for use in embodiments of the disclosed methods include the Duke University Medical Center Cyclotron and the University of Washington's Scandatronix MC-50 cyclotron. In certain embodiments, an alpha particle generator, such as a table-top generator, is used to produce an alpha particle beam. Such alpha particle generators are known to those of ordinary skill in the art. For examples of table-top alpha particle generators, see, Ji et al., Production of Various Species of Focused Ion Beam, Rev. Sci. Instrum., 2002, 73, No. 2, pp. 822-824, and Schneider, Operation of the Low-Energy Demonstration Accelerator: the Proton Injector for ATP, Proc. 1999 IEEE Particle Accelerator Conf., pp. 503-507 (IEEE Catalog No. CH36366, 1999), both of which are incorporated herein by reference. One example of a suitable table-top generator is a multicusp plasma generator employing a multicusp ion source, which was developed at Lawrence Berkeley National Laboratory in Berkeley, Calif. In principle, virtually any flux is acceptable for producing molybdenum-99, however, for commercial usage, the alpha particle beam flux typically is at least about 1016 α/(cm2)s. The alpha particle beam flux also may be less than about 1018 α/(cm2)s. For efficient transmutation of zirconium-96, in certain embodiments of the disclosed methods, the alpha particle beam has an energy of from about 10 to about 50 MeV, such as from about 10 to about 30 MeV. In certain embodiments, the alpha particle beam has an energy distribution centered at from about 10 to about 30 MeV, for example at about 14 MeV or about 15 MeV. In some cases, the beam can be substantially monoenergetic having a beam energy of about 14 MeV or about 15 MeV. A lower energy may be useful for reducing production costs. A person of ordinary skill in the art will be able to arrive at an optimum beam energy for a particular application by alpha transport modeling. Input parameters include the cross sections for the starting material, the initially-present impurities and the impurities formed by irradiation. The optimum beam energy can be higher or lower than the peak beam energy. Although other flux levels may be used, when an alpha particle beam has a flux of about 1016 α/(cm2)s and a beam energy of about 14 MeV, the target may be irradiated for a period of from about 2 hours to about 20 hours. In the production of molybdenum-99, such a target may have a zirconium-containing material thickness relative to a beam vector, for example, of about 100 μm or about 200 μm. Specific irradiation times, energy levels and flux levels can be selected by a person of ordinary skill in the art upon consideration of target system parameters, the target thickness and the desired product specifications. Because alpha particles have a relatively short path length (of the order of about 200 μm) within the target material, the target can be relatively thin. With reference to FIGS. 3A and 3B, the disk 70 may have a thickness of starting material, such as Zr-containing material, of less than about 0.5 mm, and in some embodiments, a thickness of less than about 200 μm or even 100 μm. Such relatively thin targets generally effectively dissipate heat. Because only the outer few microns of the target may be effectively irradiated with alpha particles in certain embodiments of the disclosed method, only the first few atomic layers of the target need to be removed following irradiation. Sputtering processes, such as focused ion beam mediated sputtering, are particularly useful for harvesting small amounts of irradiated material from the target. Irradiated material also can be removed by dissolving or contacting the irradiated material with a reagent. For example, in one embodiment, the irradiated area can be contacted with a reagent, such as aqua regia, to remove a portion of the target. Simple mechanical techniques also can be used to remove irradiated material from the target, as is known to those of ordinary skill in the art. For example, abrasion and/or mechanical skimming can be used to remove a portion of target material. Following removal of the irradiated material, the newly exposed target surface can be subjected to alpha particle irradiation and the process repeated to provide a continuous process. Methods for isolating the product radioisotope from irradiated target and starting materials are described herein. These methods can be used alone or in combination to provide the desired compositions including the product, such as molybdenum-99. In certain embodiments of the disclosed methods for the production of molybdenum-99, only the molybdenum-99 needs to be purified from the zirconium-containing target or starting material, because the irradiation process generally does not produce other products. Thus, in these embodiments, the crude mixture produced by irradiation primarily includes zirconium and molybdenum-99, and is relatively tractable compared to the mixture produced by the fission-based process. One embodiment of the disclosed methods for purifying molybdenum-99 exploits the different solubilities of molybdenum and zirconium species in alkaline solution. This solubility-based method is compatible with the zirconium-based starting materials described above with respect to the target. One embodiment of this method includes the following basic steps: (1) contacting the irradiated target with aqua regia; (2) evaporating the aqua regia solution to dryness to yield a residue including molybdenum and zirconium species; (3) contacting the residue with an alkaline solution; and (4) separating the alkaline solution from the insoluble material. The molybdate salts are highly soluble at elevated pH, whereas zirconium oxides and hydroxides are not. In general, any pH above about 2 provides sufficient solubility differences to separate molybdenum compounds from zirconium compounds. In one embodiment, solutions having a pH of greater than about 6 can be used to separate molybdenum from zirconium. In such embodiments, sodium hydroxide solutions having a molarity of from about 0.2 to about 0.3 (or pH values of about 13.3-13.5) can be used for solubility-based separation. This method provides a purified solution containing molybdenum-99. The method also can be repeated one or more times, such as from one to five times, to increase the molybdenum purity. Typically, a single purification according to this protocol provides a composition including molybdenum-99 and containing less than about 1% zirconium. In certain embodiments, a single solubility-based purification procedure provides molybdenum-99 of analytical purity. Some embodiments of the disclosed methods for purifying molybdenum-99 from irradiated target material involve ion-exchange chromatography. Molybdenum and zirconium species can be resolved according to embodiments of this method using either strongly basic or weakly basic ion-exchange resin. Suitable ion-exchange resins are well known to those of ordinary skill in the art and are commercially available. In general, the zirconium species exhibit a low affinity for anion-exchange resins until the acid concentration is greater than about 8 M. In contrast, molybdate exhibits a strong affinity for anion-exchange resins above an acid concentration of about 4 M, but a much lower affinity below this concentration. This acid-concentration dependent difference in affinity provides the basis for successful resolution of zirconium/molybdenum mixtures. The ion-exchange purification can be repeated to form molybdenum-99 of increased purity. However, in certain embodiments, a single ion-exchange purification provides a decrease of from about two to about three orders of magnitude in zirconium. Typically, from one to five repetitions of the ion-exchange protocol results in molybdenum-99 of analytical purity. In one embodiment of a disclosed purification method, the ion-exchange protocol is performed in combination with the solubility-based protocol described above to provide purified molybdenum-99. In this tandem protocol, the solubility-based purification described above can be performed before or after the ion-exchange protocol. Some embodiments of the disclosed methods include purifying molybdenum-99 from targets including zirconium species by fluorination. These embodiments take advantage of the different vapor pressures of molybdenum fluoride and zirconium fluoride species. For example, molybdenum pentafluoride (MoF5) has a boiling point of 213° C. and molybdenum hexafluoride (MoF6) has a boiling point of 34° C., whereas zirconium fluorides are non-volatile at these temperatures. In these embodiments, the irradiated target can first be subjected to exhaustive fluorination using a fluorinating agent. The resulting molybdenum fluoride can comprise, for example, about 80% or greater molybdenum-99. Any fluorinating agent can be used, including, without limitation hydrogen fluoride, nitrogen trifluoride, fluorine gas or combinations thereof. Other fluorinating agents can be substituted for hydrogen fluoride, nitrogen fluoride and fluorine, as is well known to those of ordinary skill in the art. The fluorinating agent can optionally be delivered in combination with a carrier, for example a carrier gas (such as helium, argon and the like) or a carrier solvent (such as water, fluorinated hydrocarbons or the like). As is known to those of ordinary skill in the art, solvents, including water and acetonitrile, can be used to modify the reactivity of the fluorinating agent. In embodiments of the disclosed methods that use a fluorinating agent, the fluorinating agent can be activated either in situ or prior to contact with the irradiated target material. Any suitable activation method can be used, such as microwave activation, which generates free radicals via photon-induced homolysis of a bond. Because molybdenum fluoride materials are relatively volatile, in comparison to zirconium fluoride materials, the desired molybdenum fluoride species can be isolated from the zirconium-based materials. An inert carrier gas, such as helium, a fluorocarbon (e.g., Freon) and/or argon can be used to flush away a molybdenum fluoride compound or compounds that are collected via, for example, a low-temperature trap. Suitable apparatuses for performing the disclosed separation procedures are well known to those of ordinary skill in the art. A system for performing a fluorination process can include, for example, a filter positioned between a fluorination chamber and the low-temperature trap, e.g., below ambient temperature. This filter can be used to prevent solid materials, such as non-volatile zirconium species, from being swept into the low-temperature trap. When the molybdenum fluoride or fluorides have been isolated, they can be converted to molybdate salts via hydrolysis with an alkaline solution, such as a sodium hydroxide solution. The resulting material, a molybdate solution, can optionally be subjected to further purification, such as by ion exchange. Embodiments of molybdenum-99 compositions produced according to certain variations of the disclosed methods may be substantially free of impurities that typically accompany fission-produced molybdenum-99 compositions. For example, fission-produced molybdenum-99 compositions can include one or more gamma particle emitters, such as iodine-131 (at levels of about 1.46×10−7), iodine-132 (at levels of about 3.0×10−5), ruthenium-103 (at levels of about 3.0×10−5 or about 1.6×10−7) and/or tellurium-132. Beta emitters, such as strontium-89 and strontium-90 also can be included in molybdenum-99 compositions produced by a fission process (in amounts as high as 5×10−6) In certain embodiments, the disclosed molybdenum-99 compositions include other isotopes, such as molybdenum-93, molybdenum-93m, niobium-96, niobium-95, niobium-92, and/or strontium-89. In certain embodiments of the molybdenum-99 compositions disclosed herein, the compositions have a higher activity concentration than molybdenum-99 compositions produced using a fission-based process. Pure molybdenum-99 has a specific activity (SPA) of 4.8×105 Ci/g. A fission reactor molybdenum-99 composition has a maximum SPA of 9.6×104 Ci/g at discharge and rapidly reduces to 1.6×104 Ci/g for a “7 day” molybdenum-99 composition. When 100% zirconium-96 material or even 70% zirconium-96-containing material is irradiated, as described above, the resulting molybdenum-99 composition can have a concentration of molybdenum-99 as much as four times greater than that of a fission reactor molybdenum-99 composition. Certain embodiments of the disclosed molybdenum-99 compositions have an activity concentration or an SPA value of at least about 19.2×104 Ci/g at discharge. For example, some embodiments of the disclosed molybdenum-99 compositions have an activity concentration or an SPA value of at least about 38.4×104 Ci/g at discharge. The SPA values of a molybdenum-99 composition can be determined by calculating isotope cross sections using the Empire II Computer software code, available from, for example, The Nuclear Energy Agency, EMPIRE-II 2.18, Comprehensive Nuclear Model Code, Nucleons, Ions Induced Cross-Sections, at, for example, http://www.nea.fr/abs/html/iaea1169.html (Aug. 2, 2004), which is incorporated herein by reference. Using the initial enrichment amounts (e.g., 70% zirconium-96) for a given starting material, the isotope production value is calculated using an isotope generation and depletion code system, such as ORIGEN (a well-published code) developed for the Nuclear Regulatory Commission and the Department of Energy to satisfy a need for an easy-to-use standardized method of isotope depletion/decay analysis for spent fuel, fissile material, and radioactive material. ORIGEN computer software code solves equations of radioactive growth and decay allowing continuous first order chemical processing and a neutron flux described by a three-region spectrum. Complex decay and transmutation schemes can be treated. ORIGEN code is available from, for example, The Nuclear Energy Agency at http://www.nea.fr/welcome.html or from the Radiation Safety Information Computational Center (RSICC), at http://www-rsicc.ornl.gov/rsicc.html or http://www.ornl.gov/sci/origen-arp/origen-arp.html, all of which are incorporated herein by reference. The output from the isotope production code provides the mass (in grams) of all of the molybdenum isotopes produced. From this, the SPA can be calculated by multiplying 4.8×105 Ci/g by the mass of the molybdenum-99 divided by the mass of all the molybdenum isotopes including molybdenum-99. Molybdenum-99 compositions produced as described above can be used to generate technetium-99m. In one embodiment of a method for producing technetium-99m, a chromatographic generator column is charged with an alumina adsorbent. The adsorbent is then equilibrated using a salt solution, such as an ammonium nitrate or saline solution. Particular examples use 0.1 M NH4NO3 for column equilibration. Molybdenum is loaded on the column as a MoO42− or hydrated MoO3 solution, typically at a pH of from about 3 to about 4. This loading solution can be prepared, for example, by titrating about 1 mL of a 0.003 mg/L molybdenum-99 stock solution (ca. 3 mg) with 1 M HNO3. Technetium-99m is eluted from the loaded column using a salt solution, such as, without limitation, 0.1 M NH4NO3, normal saline or both. The eluted technetium-99m solution may be used without further purification. However, in certain embodiments, the technetium-99m solution can be further purified by, for example, loading onto a technetium-99m concentrator column containing an anion-exchange resin, for example, AF W 1X8, 100-200 mesh in the NO3− form (equilibrated with 0.1 M NH4NO3) or other anion-exchange resin known to those of ordinary skill in the art. The concentrator column typically is washed with a small amount of salt solution, such as 0.1 M NH4NO3, followed by a small amount of deionized water. Technetium-99m can be eluted from the column using a reductive solution, such as a solution containing a complexing agent. In one embodiment, the reductive solution is prepared using an ethylenediamine (EDA) complexing agent. In one example, the reductive solution is prepared using about 0.004 parts SnCl2, about 1 part 10% EDA/H2O, about 1 part 0.1 M NaOH and about 10 parts deionized water. Other suitable complexing agents for eluting technetium-99 include molecules containing at least one amine, amide, ketone, carboxy, and/or sulfhydryl moiety. Particularly useful complexing agents are chelating agents that include at least two of these moieties. As is known to those of ordinary skill in the art, the complexing agent, such as EDA, can be exchanged for another ligand. For example, upon acidification of a technetium-99m complex to a pH of about 4, the EDA ligand exchanges with other ligands, such as citrate or gluconate. Citrate and gluconate are typical ligands used in processes for labeling tissue specific targeting agents with technetium-99m. The foregoing disclosure is further explained by the following non-limiting examples. Unless indicated otherwise, parts are parts by weight, temperature is given in Celsius or is at room temperature and pressure is at or near atmospheric. This example describes the irradiation of a target comprising zirconium to produce a composition comprising molybdenum-99. A target of zirconium metal in the shape of a disk about one inch in diameter and about three millimeters thick was irradiated with an alpha particle beam having an energy of about 28 MeV using the University of Washington's Scandatronix MC-50 cyclotron for ca. 15 minutes. Analysis of the gamma spectrum of the irradiated target shows the presence of molybdenum-99, molybdenum-93m, niobium-96, niobium-95, niobium-92m, and strontium-89. This example demonstrates a method for isolating molybdenum-99 from an irradiated zirconium target. This exemplary method includes the following steps: (1) dissolving the irradiated target in aqua regia; (2) evaporating the aqua regia solution to leave a residue; (3) treating the residue with an alkaline solution; and (4) separating the soluble molybdenum species from the insoluble zirconium-containing material. To demonstrate the effectiveness of this method, a 0.5333 g zirconium foil (99.7% purity, 0.2 mm thickness) and 0.0947 g MoO3 were combined and treated with aqua regia (3:1 volume/volume solution of HCl/HNO3 total volume of 6 mL). The mixture was heated gently for approximately 30 minutes to facilitate dissolution. A portion of the resulting solution—containing 0.048 grams of zirconium and 0.0032 grams of molybdenum—was placed in a glass beaker and evaporated to dryness using a hot plate. The residue was suspended in 5 mL of 0.25 M NaOH and heated to boiling for 20 minutes. After cooling, the suspension was centrifuged and the supernatant was analyzed using ICP-AES (inductively-coupled plasma-atomic emission spectroscopy). This analysis indicated that the supernatant had a molybdenum concentration of 0.45 g/L, which represented nearly 100% recovery of molybdenum. The zirconium concentration of the supernatant was less than 0.012 g/L. Thus, this example demonstrates that molybdenum-99 can be effectively separated from zirconium-containing materials by exploiting the different solubilities of molybdate salts and zirconium species in alkaline solutions. This example describes a method for purifying molybdenum-99 from zirconium-containing materials via ion-exchange chromatography. The irradiated target is dissolved in a suitable medium, such as aqua regia. The resulting solution is adjusted to be approximately 4 M in chloride. The solution is passed through an anion-exchange column and the column is washed with 4 M HCl to remove nitrate and residual zirconium. The desired molybdenum-99 product is eluted from the column using a dilute acid solution. The efficacy of this purification method was demonstrated using a solution comprising both zirconium and molybdenum species. A 0.5333 gram zirconium foil (99.7% purity, 0.2 mm thickness) and 0.0947 gram MoO3 were combined and treated with aqua regia (3:1 volume/volume solution of HCl/HNO3 total volume of 6 mL). The mixture was heated gently for approximately 30 minutes to facilitate dissolution. The resulting solution was diluted with deionized water from its initial chloride concentration of 9 M to a chloride concentration of 4 M and a concentration of 10.4 mM Mo and 134 mM Zr. 11 mL of the diluted solution was applied to a column of strongly basic anion-exchange resin (BioRad AG1-X4, 50-100 mesh 13 grams, ca. 40 mL) and passed through the column at a rate of about 0.5 mL/minute. Then, about 50 mL of 4 M HCl (“scrub” solution) was passed through the column, followed by 60 mL of 0.04 M HCl (“strip” solution). The results of two trials performed according to this protocol are recorded in Table 1. TABLE 1Eluted FeedEluted ScrubEluted StripSolutionSolutionSolution(% Recovery)(% Recovery)(% Recovery)Trial 1:Mo0.180.070.12Zr1021.70.024Trial 2:Mo0.10.080.01Zr94130.07 The results recorded in Table 1 demonstrate that Zr species can be separated from Mo species using ion-exchange chromatography. Several additional stripping conditions also were evaluated to improve the recovery of Mo from the ion-exchange resin. To evaluate Mo recovery conditions, an analyte solution (0.014 M molybdate) was prepared by dissolving ammonium molybdate (88 mg) in HCl (4 M; 50 mL). Each anion-exchange column was prepared using either strongly basic BioRad AG1X4 (ca. 1 gram per column, 50-100 mesh, Cl− form, 5.5 meq reported capacity per gram of dry resin) or weakly basic BioRad AG3X4 (ca. 1.25 grams per column, 100-200 mesh, 5.5 meq reported capacity per gram of dry resin). For each recovery assay, the resin was conditioned by passing several column volumes of feed solution (4 M HCl) through the resin. The analyte solution (1.5 mL) was added, followed by a feed solution (HCl 4 M). Following the addition of 10-15 mL of 4 M HCl, about 10-15 mL of deionized water (scrub) was added to the column, followed by 10-15 mL of the “strip” solution. The feed, scrub and strip effluents were separately collected and, following acidification by the addition of an equal volume of HCl (6M), were analyzed for sodium and molybdenum by ICP-AES. The results of this analysis for assays using the strongly basic AG1X4 resin are recorded in Table 2. TABLE 2% Mo in% Mo in% Mo inFeedScrubStripTotal % MoStrip SolutionEffluentEffluentEffluentRecovery0.25 M NaOH0.63.32.76.62.5 M NaOH0.081.472731 M NaOH0.33.435383:1 (v:v) 0.250.30.26767M NaOH/30%H2O2 solution The results recorded in Table 2 demonstrate that strongly basic anion-exchange resin retains molybdate from a 4 M HCl solution with high affinity. Moreover, hydroxide solutions or mixed NaOH/H2O2 solutions can be used successfully to recover molybdate during anion-exchange chromatography. Weakly basic anion-exchange resin also can be used to separate zirconium from molybdenum. For example, BioRad AG3X4 weakly basic anion-exchange resin (10 grams, 100-200 mesh, 5.5 meq reported capacity per gram of dry resin) was loaded onto a 1.5 cm internal diameter column having, after loading, an approximately 25 mL dead volume. The column was equilibrated using 0.04 M HCl (ca. 25 mL) followed by 4 M HCl (ca. 20 mL). A diluted aqua regia solution containing molybdenum and zirconium (110.5 mL, prepared as described above) was added, followed by 50 mL of 4 M HCl, 20 mL deionized water, and 50 mL of 3 M NaOH. The results of this trial are recorded in Table 3. TABLE 3% in Feed% in Scrub% in StripTotal %EffluentEffluentEffluentRecoveredMo0.39145064Zr937.40.32101 The Mo/Zr separation factor achieved in the above example is about 160. Thus, the results recorded in Table 3 demonstrate that weakly basic anion-exchange resin can be used to separate molybdenum species from zirconium species via anion-exchange chromatography. This example describes the production of 99Mo via the alpha-particle bombardment of a natural zirconium target. Copper (5N purity) and zirconium (3N purity) disks 2 inches in diameter and 0.125 inches thick were purchased from Electronic Space Products International (ESPI, Ashland, Oreg.) and were milled to 1 inch diameter. The disks were washed with a detergent solution to remove surface impurities, rinsed with tap water, and dried in air before use. All bombardments were performed at the alpha accelerator at the University of Washington Department of Radiation Oncology using a stream (10 μA) of alpha particles of energy ˜28 MeV. The metal disks (three copper, ˜14.64 g; three zirconium, ˜11.2 g) were placed individually in the sample holder of the accelerator so that the alpha beam was completely absorbed by the metal target. The copper and zirconium targets were bombarded for 2-3 and 10-15 minutes, respectively. After irradiation, the targets were stored for approximately two days before being placed on an apparatus for gamma-energy analysis. The samples were counted at about 15 cm from the face of an intrinsic germanium detector. The analyses were performed using a Digital Spectrum Analyzer (Canberra) that had been setup and tested to correct accurately for system dead time for count rates much higher than those produced by the irradiated targets. The detector was calibrated with NIST-traceable standards with typical total propagated uncertainties of 1.5% at 1-sigma. The quoted uncertainties included all known sources of error including the counting statistics, calibration, and uncertainties in the half-lives and gamma intensities. Daily control counts were performed with 241Am, 60Co and 137Cs standards to verify the continuing detector calibration. All nuclear data were taken from tables (Browne and Firestone, 1986). 67Cu and 67Ga both have some of the same gamma ray emissions, but their relative gamma intensities are quite different. A simultaneous fit using the potential gamma emissions from both isotopes indicated virtually 100% of the gammas were emitted from 67Ga with no 67Cu activity. All data were corrected to an end of bombardment time on Jul. 26, 2004, at 9:21 am PDT. The three copper targets were used as standards to verify the flux of the alpha-particle beam. Target Cu-1 was irradiated for 2 min; Cu-2 and Cu-3, for 3 min. The activities (Bq) of the radioisotopes that were produced during the relatively brief bombardments are listed in Table 4. TABLE 4RadioisotopeHalf-lifeCu-1 (Bq)Cu-2 (Bq)Cu-3 (Bq)Zn-65244.06d2.60 × 104 ± 2.3%3.77 × 104 ± 2.3%3.92 × 104 ± 2.3%Ga-669.49h1.17 × 107 ± 2.9%1.65 × 107 ± 2.5%1.72 × 107 ± 2.5%Cu-6761.83h<1.5 × 103<1.5 × 103<1.5 × 103Ga-673.2612d8.99 × 105 ± 2.0%1.36 × 106 ± 2.0%1.40 × 106 ± 2.0% The flux of the alpha-particle beam was calculated using the values for the quantities of radioisotopes produced, the known cross-sections for the (α,x) reactions on copper, and the time from end of bombardment. This flux then was used to calculate the cross-sections for (α,x) reactions on zirconium. The three targets of natural zirconium were irradiated to provide data for determining the cross-section for 99Mo production. Zirconium targets Zr-1 and Zr-3 were irradiated for 15 min; Zr-2, for 10 min. Table 5 lists the activities (Bq) of the radioactive isotopes that were produced directly after the end of bombardment. TABLE 5RadioisotopeHalf-lifeZr-1 (Bq)Zr-2 (Bq)Zr-3 (Bq)Sr-8950.53d3.14 × 104 ± 3.0%2.01 × 104 ± 2.6%3.18 × 104 ± 2.5%Nb-92m10.15d6.88 × 104 ± 2.3%4.74 × 104 ± 2.6%7.25 × 104 ± 2.2%Mo-93m6.85h2.04 × 107 ± 3.5%1.37 × 107 ± 3.6%2.12 × 107 ± 3.6%Nb-9534.991d1.05 × 104 ± 2.7%6.96 × 103 ± 3.5%1.05 × 104 ± 2.8%Nb-9623.35h2.03 × 105 ± 1.6%1.32 × 105 ± 1.4%2.01 × 105 ± 1.5%Mo-9965.94h1.37 × 105 ± 1.9%8.73 × 104 ± 2.1%1.35 × 105 ± 1.9% Natural zirconium can include 90Zr (51.45%), 91Zr (11.22), 92Zr (17.15), 94Zr (17.38) and 96Zr (2.80). Considering the reactions that can occur during bombardment with alpha particles, the observed radioactive products were primarily as follows: 90Zr (α,d) 92mNb 90Zr (α,n) 93mMo 91Zr (α, α2p) 89Sr 92Zr (α,2n) 93mMo 92Zr (α,p) 95Nb 94Zr (α,d) 96Nb 96Zr (α,n) 99MoThe stable Mo products were primarily produced by the following reactions: 90Zr (α,2n) 92Mo 91Zr (α,n) 94Mo 92Zr (α,n) 95Mo 92Zr (α,2n) 94Mo 94Zr (α,n) 97Mo 94Zr (α,2n) 96Mo 96Zr (α,2n) 98Mo By way of theory, many of the short-lived products also may have been formed because the beta-gamma dose rate of the targets decreased very quickly after the end of bombardment. However, the identities of these products could not be determined because the targets were allowed to cool for approximately two days before the gamma-energy analysis. FIG. 4 shows the results from three measurements of the cross section for the 96Zr (α,n) 99Mo reaction. The previous experimental results (Chowdhury et al., 1995), represented by diamonds 90, indicated that the cross section is at least 120 mb with a peak near 15 MeV. Calculations performed using the EMPIRE-3 code, represented by line 100, indicated a cross-section value greater than 180 mb at about 14 MeV. The one-group thick-target cross section, represented by line 110, had an average effective cross-section of about 66 mb over the energy range 0-30 MeV. The one-group thick-target cross section was derived as follows. The basic equations for the production (activity) of a radioactive isotope from a stable target for a given constant flux (φ) and single cross section value (σ1) are: N 2 ⁡ ( t ) = N 1 ⁡ ( 0 ) ⁢ φσ 1 λ 2 ⁢ ( ⅇ - φσ 1 ⁢ t - ⅇ - λ 2 ⁢ t ) ( 1 ) A 2 ⁡ ( t ) = λ 2 ⁢ N 2 ⁡ ( t ) ( 2 ) λ 2 ⁢ ln ⁢ ⁢ 2 / t 1 / 2 ( 3 ) where N2(t) is the number of atoms of isotope 2 (radioactive); N1(0), the initial number of atoms of isotope 1 (stable); φ, the total particle flux; σ1, the one-group effective capture cross section for isotope 1; λ2, the decay constant for isotope 2; t, the irradiation time for production; t1/2, the half life of isotope 2; and A2(t), the activity of isotope 2 at time t. Note that the above equations assume that the irradiation time is short enough such that there is no appreciable burnout of the target (isotope 1). For very short irradiation times (t<<t1/2 and t<<1.0/φσ1) these equations give:A2(t)=N1(0)φσ1λ2t (4) These same equations apply to situations involving one-group thick-target cross sections. First, they describe the thick target result, where σ1 corresponds to the effective one-group cross section (σ1) for a thick target and φ, N1(0), and A2 are the average flux (φav), the number of target atoms, and the resulting activity for the whole thick target, respectively. Second, they give the activity produced (dA2) in a “thin” slice (dx) of the thick target (FIG. 5) with a face perpendicular to the alpha beam and the dimensions of area (Ar) and length [range of alphas for incident energy at front face (Ra)] at a position along the x-axis (x), where the energy of the alpha particle is E(x) with a corresponding cross section σ1 [E(x)]. The mass (dm) of the thin slice for very short irradiation times (t<<t1/2 and t<<1.0/φσ1) is then:dm=ρArdx (5)where ρ is the density (g/cc) of the thick target. From this the activity in the thin slice is: dA 2 = dN 1 ⁡ ( 0 ) ⁢ φ ⁡ ( x ) ⁢ σ 1 ⁡ ( x ) ⁢ λ 2 ⁢ t ( 6 ) dN 1 ⁡ ( 0 ) = d ⁢ ⁢ m ⁡ ( N 0 / Awt ) ( 7 ) where N0 is Avogadro's number (6.02×1023) and Awt is the atomic weight of the target. The flux φ(x) is assumed to be constant and given as: φ ⁡ ( x ) = I 0 / Ar ( 8 ) where Io is the incident current (α/s) of α-particles crossing the cylinder surface (area Ar) per unit time.Combining Eqs. (5) to (8) then gives: dA 2 = ( N 0 ⁢ I 0 ⁢ ρ Awt ) ⁢ σ 1 ⁡ ( x ) ⁢ λ 2 ⁢ tdx ( 9 ) Integrating Equation (9). A 2 = N 0 ⁢ I 0 ⁢ ρλ 2 ⁢ t Awt ⁢ ∫ 0 Ra ⁢ σ ⁡ ( E ⁡ ( x ) ) ⁢ ⅆ x ( 10 ) Given a “microscopic” (energy-dependent) cross section and the energy dependence E(x) as the alpha particle slows down in the thick target, Eq. (10) can be used to calculate the “total” activity produced for a given σ(E(x)). Using Equation (4) in terms of the one-group cross section results in: A 2 = N 0 ⁢ I 0 ⁢ λ 2 ⁢ t ⁢ ⁢ ρ Awt ⁢ σ 1 ⁢ Ra ( 11 ) Hence: σ 1 = ∫ 0 Ra ⁢ σ ⁡ ( E ⁡ ( x ) ) ⁢ ⅆ x / Ra ( 12 ) As indicated above, in addition to the radioisotopes 93mMo and 99Mo, irradiation of natural (or enriched) Zr will produce stable Mo isotopes. The stable isotopes of Mo are 92Mo, 94Mo, 95Mo, 96Mo, 97Mo, 98Mo, and 100Mo. The amount of these stable isotopes has a direct influence on the specific activity of the 99Mo product. Fission of 235U targets also produces stable isotopes of Mo so that the specific activity of fission 99Mo (4.3×104 Ci/g at discharge) is significantly less than the theoretical value for pure 99Mo (4.8×105 Ci/g). The specific activity of 99Mo produced via the (α,n) reaction was compared with that of fission 99Mo. The result of this comparison provides a clear indication of whether 99Mo from the (α,n) reaction will behave similarly to fission 99Mo in existing 99Mo-99mTc column generators. Q values were computed for the reactions leading to Mo isotopes with the assumption that enriched 96Zr containing 92Zr (3%), 94Zr (27%), and 96Zr (70%) is available. The yields of both stable (95Mo, 97Mo, 98Mo, and 100Mo) and radioactive 99Mo isotopes from the (α,n) reaction with monoenergetic alpha particles of 11.0 and 15.0 MeV then were calculated. The calculated specific activities of the 99Mo product were 4.75×105 and 2.2×105 Ci/g, respectively. The value at 11.0 MeV was greater than that at 15.0 MeV because the cross sections for production of stable Mo isotopes are much lower at the lower energy. In any case, these specific activities were 5-10 times greater than that for fission 99Mo, indicating that unfavorable effects due to a large mass of Mo should not be a factor in using this 99Mo in the existing column generators. Bombardment with alpha particles of a zirconium target including a significant amount of 96Zr is a viable method for producing 99Mo that can be used in 99Mo-99mTc generators. Since natural molybdenum includes 92Mo, 94Mo, 95Mo, 96Mo, 97Mo, 98Mo and 100Mo, the presence of any of these stable isotopes in the 99Mo product may reduce its specific activity. Thus, it is desirable to enrich the zirconium target as high as possible in 96Zr. This serves the purpose of decreasing the production of 93mMo and stable molybdenum isotopes, both of which will compete with 99Mo for binding sites on the alumina column in the 99mTc generator. Throughout this disclosure, the production of 99Mo is described in detail. The processes described above also can be applied to the production of other medically useful radioisotopes. This example describes the production of some of these additional radioisotopes. Table 6 includes a list of medically useful radioisotopes that can be produced by alpha-particle bombardment of various target materials. TABLE 6NaturalPeak CSEnergyProductHalf-lifeTargetAbundance %(mb)(MeV)123I13.2h121Sb57.20%127829124I4.18d121Sb57.20%37218125I59.4d123Sb42.80%125028111In2.80d109Ag48.20%11432767Cu2.58d64Ni 0.93%80.720201Tl3.04d198Hg 9.97%0.9392918F1.83h16O99.80%1243318F1.83h15N 0.33%4301551Cr27.7d48Ti73.70%6941466Ga9.5h63Cu 69.2%5801767Ga3.26d64Zn48.00%4712067Ga3.26d65Cu30.80%9002768Ge270d66Zn27.90%5513182Sr25.4d80Kr 2.28%86929109Cd461d107Ag51.80%6.630211At7.21h209Bi 100%10803199Mo2.75d96Zr 2.80%18514186Re3.72d183W14.30%0.42430177Lu6.65d174Yb31.80%0.09525103Pd17.0d100Ru12.60%54218132Cs6.48d129Xe26.40%3.125153Sm1.93d150Nd 5.60%14.916195mPt4.02d192Os40.80%26.52147Sc3.35d44Ca 2.09%11017131I8.02d128Te31.70%0.31323166Ho1.12d163Dy24.90%0.1624 The production of each of these radioisotopes can be optimized by a procedure similar to the procedure described above with respect to the production of 99Mo. For example, in each case, the cross sections of the impurities can be determined along with the cross sections of the radioactive and stable products. Using these cross sections, the energy of the alpha accelerator can be tuned to optimize production of the desired products. Where necessary, an appropriate separation process (e.g., chemical or physical) can be used to isolate the desired products after irradiation. FIGS. 6, 7 and 8 show cross sections for isotopes resulting from alpha irradiation of 16O, 121Sb and 15N, respectively. These cross sections can be used to optimize the production of the desired products. In many cases it is possible to use this optimization method to produce substantially carrier free product radioisotopes. The present invention has been described with respect to certain preferred embodiments. However, the present invention should not be limited to the particular features described. Instead, the scope of the invention should be determined by the following claims.
059463643	description	The invention will now be described with reference to the following non-limiting example. EXAMPLE A particular example of practicing this invention is as follows. Sintering uranium dioxide nuclear fuel pellets, in the form of right circular cylinders and containing 0.25 wt % of alumina/silica in the ratio of 0.4 Al.sub.2 O.sub.3 /0.6 SiO.sub.2 are prepared by blending alumina and silica powders in a weight ratio of 0.4 Al.sub.2 O.sub.3 /0.6 SiO.sub.2 with uranium dioxide powder to achieve a total addition of 0.25 wt % of the alumina/silica with 99.75% uranium dioxide. The blended powders are dry-pressed to a green density of approximately 5.6 gm/cm.sup.3 to form powder compacts in the form of right circular cylinders for sintering to fuel pellets. The dry pressed pellets are sintered using a furnace feed gas of 75% hydrogen -25% nitrogen which has been moisturized by passing the gas through a water bubbler with the temperature of the water in the bubbler maintained at 55.degree. C. and a total furnace gas pressure of 1 atmosphere (760 mm Hg). At 55.degree. C., the vapor pressure of water is 118 mm Hg, the hydrogen and nitrogen gas pressures of the furnace feed gas are 481.5 and 160.5 mm Hg, respectively, and the H.sub.2 O to H.sub.2 ratio of the furnace gas atmosphere is 118/481.5=0.245. The sintering furnace temperature profile is maintained to provide prolonged (.about.4 hours) sintering at 1750.degree. C. in the hot or working zone of the sintering furnace. At that sintering temperature, for the H.sub.2 O to H.sub.2 ratio noted above, the oxygen free energy in the hot zone of the sintering furnace is maintained at about -70 kcal/mole, the O/U ratio of the uranium oxide during the sintering operation is maintained at about 2.005, and the vapor pressure of SiO is maintained at about 10.sup.-5 (0.00001) atmospheres. For these sintering conditions, the desired final fuel pellet density of 10.5 gm/cm.sup.3 is achieved, and the aluminum and silicon content of the final sintered pellets are within acceptable ranges of the initial amount added. The sintered pellets are tested for resistance to density changes by maintaining them in the temperature range 1700-1725.degree. C. with a furnace feed gas of 75% hydrogen-25% nitrogen which has been moisturized by passing the gas through a water bubbler with the temperature of the water in the bubbler maintained at 55.degree. C. and a total furnace gas pressure of 1 atmosphere (760 mm Hg). At 55.degree. C., the vapor pressure of water is 118 mm Hg, the hydrogen and nitrogen gas pressures of the furnace feed gas are 481.5 and 160.5 mm Hg, respectively, and the H.sub.2 O to H.sub.2 ratio of the furnace gas atmosphere is 118/481.5=0.245. At that test temperature, for the H.sub.2 O to H.sub.2 ratio noted above, the oxygen free energy in the hot zone of the testing furnace is maintained at about -70 kcal/mole, the O/U ratio of the uranium oxide during the sintering operation is maintained at about 2.005, and the vapor pressure of SiO is maintained at about 10.sup.-5 (0.00001) atmospheres. During cooling, in this test atmosphere, the O/U ratio of the uranium oxide re-equilibrates to near 2.000. For these test conditions, the desired final fuel pellet the weight loss of the pellets as a result of SiO loss is negligible, and the density changes are essentially only a result of pore volume elimination from continuation of the sintering process. While the invention has been described in connection with what is presently considered to be the most practical and preferred embodiment, it is to be understood that the invention is not to be limited to the disclosed embodiment, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims.
044217161	abstract	An apparatus which monitors a subset of control panel inputs in a nuclear reactor power plant, the subset being those indicators of plant status which are of a critical nature during an unusual event. A display (10) is provided for displaying primary information (14) as to whether the core is covered and likely to remain covered, including information as to the status of subsystems needed to cool the core and maintain core integrity. Secondary display information (18, 20) is provided which can be viewed selectively for more detailed information when an abnormal condition occurs. The primary display information has messages (24) for prompting an operator as to which one of a number of pushbuttons (16) to press to bring up the appropriate secondary display (18, 20). The apparatus utilizes a thermal-hydraulic analysis to more accurately determine key parameters (such as water level) from other measured parameters, such as power, pressure, and flow rate.
	description	1. Field of the Invention The present invention concerns an apparatus for spatial modulation of an x-ray beam, of the type having a number of planar attenuation elements for x-ray radiation that are disposed in a grid-like manner on a carrier and that, independently of one another, can be pivoted or tilted piezoelectrically between at least two positions. The invention furthermore concerns an x-ray image system with such a modulation apparatus as well as different methods for operation of such an apparatus. 2. Description of the Prior Art Laminar x-ray image systems are used primarily in medical diagnostics in order to acquire radiographic images of the inside of the body of a patient. The patient is penetrated by an x-ray field extending perpendicularly in two dimensions to the propagation direction, and the spatially dependent attenuation of the x-ray radiation received behind the patient is represented or evaluated as image information. In addition to conventional radiography, laminar x-ray imaging systems are used in fluoroscopy as well as, more recently, in so-called multi-slice systems in computed tomography. The radiation dose to which the patient as well as the medical personnel is exposed during the examination plays a significant role in applications in medical x-ray diagnostics. A reduction of the applied x-ray dose can be achieved by use of a semi-transparent pre-filter that has a central opening for the unattenuated passage of x-ray radiation therethrough. By suitable placement of such a filter as is known, for example, from U.S. Pat. No. 5,278,887, only the region of the patient within the two-dimensional radiation field is charged with the necessary dose that is of interest for the user of the x-ray image system. The regions in the image lying outside of this ROI (region of interest) are nevertheless recognizable, albeit with reduced contrast. This technique in fact effects a significant dose reduction in the border regions of the image, but can be adapted only with difficulty to different subject shapes and sizes. Even with the use of such a filter technique, the dose at specific regions of the body to be examined is locally higher by multiple times more than would be necessary for a good contrast. This problem particularly occurs in body regions in which regions of much stronger x-ray absorption and regions of much weaker x-ray absorption lie next to one another. Since the diagnosing physician must normally examine all organs of an x-ray image, the applied x-ray dose is set such that a sufficient signal-to-noise ratio is achieved for all objects acquired in the image. Apparatuses for spatial modulation of the radiation field that are positioned between the x-ray source and the patient are known in the field of x-ray imaging in which one-dimensional radiation fields are used in the form of fan-shaped x-ray beams for exposure such as, for example, in conventional computed tomography. In these apparatuses, for example, tongue-shaped attenuation elements are arranged in the form of a one-dimensional array corresponding to the one-dimensional extent of the radiation field. The attenuation elements can be controlled via separate actuators independently of one another, such that individual sections or channels of the one-dimensional radiation field can be weakened or modulated independently of one another by the introduction of the attenuation elements. Such an apparatus is known, for example, from U.S. Pat. No. 5,044,007, in which the attenuation elements are fashioned tongue-shaped and tiltable, and each can be tilted into the radiation field by its actuator. The control of the individual actuators ensues dependent on the x-ray radiation exiting from the body after irradiation of the body to be examined, relative to the respective channel that can be influenced with the attenuation element. The radiation dose necessary for a sufficient contrast can be locally reduced in this manner to the respectively necessary value, such that overall a reduced radiation exposure results for the patient. Similar apparatuses are known from U.S. Pat. Nos. 5,054,048, and 4,715,056, and European Application 0 251 407. In U.S. Pat. No. 5,054,048, the attenuation elements are designed as sliding elements that are moved into or out of the beam by a sliding mechanism with an electromechanical drive. The attenuation elements are wedge-shaped, such that different degrees of attenuation can be achieved by displacement thereof perpendicular to the beam direction. European Application 0 251 407 suggests the use of planar attenuation elements made from a piezoelectric material that can be tilted between two positions by the application of an electrical voltage. From U.S. Pat. No. 4,715,056, a further one-dimensional attenuation apparatus is known in which tiltable pivotable planar attenuation elements are formed from a piezoelectric material as flex transducers that can be bent into the beam path by the application of an electrical voltage. This document furthermore discloses the possibility of an electromechanical drive as well as drive by means of a step motor. In the design with the electromechanical drive, the position of the attenuation elements is derived from the current strength of the current flowing through the electromagnet, namely the activating current. A further apparatus for spatial modulation of a two-dimensional x-ray field is known from the German Application 102 21 634 (published after the priority date of the present application). In this apparatus, flex transducers (arranged in the form of a grid) to which self-supporting planar attenuation elements are attached, are aligned such that the attenuation elements stand in the beam direction. A minimal beam attenuation ensues in this position. By deflection of individual flex transducers via an electrical control, the attenuation at this location can be specifically increased. Knowledge of the position of each individual attenuation element during the image acquisition is necessary to damp the fluctuation behavior of the flex transducer and for image post-processing. The detection of the current position of each attenuation element ensues in this apparatus with an optical measurement arrangement that detects light passing through the grid of attenuation elements. For this purpose, a light source is necessary at the input side of the grid and a light deflection device is necessary at the output of the matrix. The respective positions of the attenuation elements are determined and evaluated by the shadowing thereof on a photodiode array caused by the attenuation elements. Errors can occur with such an optical detection due to light scattering and image blurring. Furthermore, light channels must be present in the grid mounting in order to be able to conduct the optical projection onto the photodiode array. An object of the present invention is to provide an apparatus for spatial modulation of an x-ray beam by means of planar attenuation elements arranged in a grid, which enables a precise determination of the position of every attenuation element in real time. The object is achieved by an apparatus according to the present invention having a number of planar attenuation elements for x-ray radiation disposed in a grid arrangement on a carrier, which can be pivoted or tilted piezoelectrically independently of one another between at least two positions. One or more sensors with which a piezoelectrically-caused length and/or width and/or position change of the piezoelectrically influenced regions can be detected are disposed on piezoelectrically influenced (in terms of length and/or width and/or position change) regions of the attenuation elements or piezoelectric drive elements that are connected with the attenuation elements. The background for the present invention—as well as some of the apparatuses of the prior art—is the realization that the spatial distribution of the radiation field before the passage through the body of the patient is, aside from interference effects, practically homogenous, while as a consequence of the absorption ratios of the patent body the dynamic range in the radiation field immediately before the x-ray detector can amount to 1:1000 or greater. With the present apparatus, a further dose reduction is achieved because, within the two-dimensional radiation field before the passage through the body, a dose is applied that has an intensity no higher than is sufficient for a good contrast at this location to be obtained from the x-ray radiation that strikes the x-ray detector. The present apparatus thus enables a fast, adaptive, image-content-controlled spatial radiation attenuation in the two-dimensional radiation field in front of the patient body. The achievable dose reduction is based on individually applying, via controlled pivoting or tilting of the individual attenuation elements within their respective grid region (also designated in the following as a beam channel or cell), in each image region only as much of a dose as is necessary and that location to achieve a satisfactory signal-to-noise ratio. The image signal acquired by the x-ray detector must be corrected dependent on the set spatial transparency of the apparatus operating as a radiation attenuator before further processing or display. For this, knowledge about the current position of each individual attenuation element during the image acquisition is necessary. In the present apparatus, this position information is obtained via the sensor signals that respectively detect the current flexing or expansion state of each individual attenuation element, or the piezoelectric actuation of the attenuation element, such that the exact position can be derived therefrom. This position is linked with the degree of attenuation of the respective channel, such that with this information an x-ray image acquired with the inventive apparatus can be normalized. The entire radiation distribution thus can be reverse calculated using the individual settings or deflection positions of the attenuation elements determined with the sensors. In contrast to the aforementioned method for the determination of the position via the controller, the inventive solution offers the advantage of a higher degree of precision and, with respect to the technique of the subsequently published document, additionally offers the advantage of a smaller technical expenditure. By the inventive direct detection of the attenuation element, or its actuation element, no interfering signals of other attenuation elements occur, as with optical detection. Moreover, the attenuation elements or their piezoelectric drives can be completely glued at one end into the carrier serving as a mounting. This simplifies the production and leads to a higher stability of the mounting. The present invention enables the direct detection of the current position of each attenuation element in real time without the need for elaborate optics. Modifying influences, such as fluctuations and external interferences that can be individually different for each of the attenuation elements, also can be compensated. The present apparatus can be particularly advantageously used in an x-ray image system in connection with a controller, with the position of a respective attenuation element currently detected with the sensor being compared with a reference position to be reached, and the attenuation element is controlled to reach the desired reference position. Naturally, it is not necessary for this purpose to concretely, spatially calculate the current position of the attenuation element. Rather, a known association between the sensor signal and the position of the attenuation element is sufficient, such that control can be based directly on the sensor signal. In an embodiment of the present apparatus, the attenuation elements are themselves piezoelectric flex transducers attached on one side of the carrier. These flex transducers either can be formed directly from a material that strongly absorbs x-ray radiation, or can be coated with such a material such as, for example, tungsten. In a further embodiment, the attenuation elements are formed as self-supporting elements, made of a material that strongly absorbs x-ray radiation, that are connected with piezoelectric drive elements. The piezoelectric drive elements are fashioned as piezoelectric flex transducers attached approximately parallel to the flex transducers on one side of the carrier, each having a free end to which the attenuation elements are attached. A “self-supporting attenuation element” as used herein means a component that is stable, in contrast to a thin layer, and can be arranged and moved freely in space without further support. A much larger movement range can be traversed in less time by a suitable arrangement of these self-supporting attenuation elements relative to the flex transducers. For example, planar metal rods or metal plates can be used as attenuation elements. In an embodiment, the sensors for the detection of the piezoelectrically induced length and/or width and/or position change of the piezoelectrically influenced regions are tensiometer (strain gauge) strips that are attached to the flex transducers. These tensiometer strips directly detect the piezoelectrically caused expansion and thus the curvature of the flex transducer. The tensiometer strips can hereby either be glued or directly imprinted onto the flex transducer. In a further embodiment, the sensors are directly integrated into the flex transducers. This ensues by the use of a further layer made of a piezoelectric material that is a component of the flex transducer. Such flex transducers are known, and are called trimorph flex transducers. This second layer made from a piezoelectric material is used as a sensor with which the respective current flexing of the flex transducer can be detected. A carrier made from a material that optimally slightly absorbs the x-ray radiation to be modulated is used as a carrier or mounting in the preceding embodiments. In particular material made from plastic or a metal with a low atomic number are suitable. In a further embodiment of the apparatus a substrate is used as the carrier, the substrate being penetrated by passage channels running parallel to one another or aligned to the focus of an x-ray source, with the attenuation elements being disposed in these channels. The attenuation elements are arranged such that they can be tilted or pivoted within the passage channels, such that each element completely closes (blocks) its channel in one position of the element. In this case, for each attenuation element two piezo-stack actuators can be provided as drive elements that are offset from one another on respective main surfaces of the attenuation element, and that are connected to the inner wall of the passage opening. In this embodiment, the attenuation elements can be tilted on a central axis given activation of the drive elements. The sensors are arranged on the piezo-stack actuators in order to detect their expansion. This can ensue, for with tensiometer strips. In their neutral position, in which they attenuate the x-ray radiation the least within the cell or the beam channel, the longitudinal axis of each attenuation element is aligned to the focus of the x-ray source of the x-ray image system in which the elements are used. Upon activation, these attenuation elements are tilted within their respective cells such that they occupy a larger portion of the cell cross-sectional area. Due to the grid-like arrangement of the individual attenuation elements, a grid of controllable beam channels is created. The grid does not need to be sub-divided nearly as finely as the grid of the laminar x-ray detector in the x-ray image system. Due to the proximity of the attenuation elements to the focus of the x-ray source, they are deliberately imaged out-of-focus on the x-ray detector. It is advantageous for the shadowing effect of adjacent attenuation elements to partially overlap on the x-ray detector, since in this manner a more spatially consistent shadowing is created. The control of the quantum flow of the x-ray radiation in each radiation channel ensues by the variation of the angle of inclination or tilt angle of the attenuation elements. When the attenuation element is aligned with its longitudinal axis exactly on the focus of the x-ray source, the absorption in the radiation channel is at a minimum. In this position, the maximum value of the radiation is allowed to pass in this channel. When the attenuation element is maximally pivoted or tilted, radiation attenuation ensues in a larger portion of the radiation channel. An effective width of the absorbing part of the attenuation element in the cell that is effective for the absorption of the x-ray radiation in the channel region, corresponding to 11.43 times the actual width of this absorbing part is obtained by tilting the attenuation element of 5° relative to the neutral position. Given a width of, for example, 125 μm, this yields an effective width of 1.5 mm for the attenuation of the x-ray radiation. From this, given an x-ray voltage of 50 to 80 keV, a quantum flow change >10–13 is obtained as an attenuation factor in the case of an attenuation element with an absorbing part made from tungsten. It is necessary for these elements to exhibit a high degree of radiation absorption in order to actually modulate the radiation, rather than merely harden it. The present x-ray image system with the inventive apparatus for spatial modulation of the x-ray beam includes an x-ray source and a laminar x-ray detector on opposite sides of an examination volume in a known manner. The inventive apparatus is disposed at the side of the examination volume near the x-ray source, in the beam path of the x-ray radiation. Furthermore, the x-ray image system has a controller to control the attenuation elements of the apparatus, preferably dependent on the spatial distribution of the x-ray radiation striking the x-ray detector. With this controller, the attenuation elements can be electronically controlled dependent on the locally received x-ray radiation or on the image content so that a consistent signal-to-noise ratio is attained with an optimally low x-ray dose. As needed, the attenuation elements are set to allow a reduced quantum flow by partial tilting. The contrast reduction thereby effected can be compensated in the image reproduction chain, for example by digital post-processing, with the effective action of the attenuation elements being detected with sensors in real time in each channel. In the image post-processing, for each pixel of the x-ray detector the actual amplitude value is multiplied with the previously measured attenuation factor at this pixel. This attenuation factor can also be composed of the shadowing effect of a number of attenuation elements, since this shadowing effect can be partially overlapped on the x-ray detector by the arrangement of the attenuation elements near the x-ray source. In an embodiment of the x-ray image system, the controller for the attenuation elements is integrated into a control loop in which the attenuation elements are controlled dependent on the measurement signals of the sensors to achieve the predetermined desired position. In this manner, even given scatterings or other influences individually acting on the attenuation elements or their drive elements, a reliable adjustment of the desired position can be achieved. The present apparatus can be used for different tasks in the field of x-ray imaging technology. In one application, the present application can serve for dose reduction, dynamic increase and/or improvement of the image quality in radiography exposures or DSA. In this application, the quantum flow is determined in the effective region by a two-dimensional x-ray detector (for example a solid-state detector) with fast sampling rate. The attenuation of the individual cells of the apparatus is detected during a first part of the exposure and the control unit uses this information for the adjustment of the attenuation of the individual channels. Higher intensity locations in the image receive fewer or no further quanta in the further exposure by reduction of the transparency by means of the beam attenuators in the second part of the exposure, while the attenuation elements remain set to the highest transparency at dark, low-intensity image locations. This application can be implemented in real time by means of the fast pivoting or tilting capability. In this application of the apparatus, a significantly smaller x-ray dose is applied in relatively transparent image regions. The factor of the achievable dose reduction is subject-dependent and can be more than a factor of 10 in an individual case. Given implementation of this application with the use of additional pre-scans, the image acquired with the pre-scan can be integrated into the end image, such that all applied x-ray quanta contribute to the end image. The reaction time of the individual radiation attenuators must be fast enough for this application and the sampling rate of the x-ray detector must be relatively high. Values from 100 ms up to 100 μs can be achieved as a reaction time of the radiation attenuator. The attenuation elements operate only in the in/out mode, meaning without the use of intermediate settings. A further application field of the present apparatus in medical diagnostics concerns the dose reduction, dynamic increase and/or improvement of the image quality in fluoroscopy. In this application, the transparency of the preceding images is used as a basis for the adjustment of the individual attenuation elements of the apparatus. Since the image content of successive images for the most part differs only a little in fluoroscopy, relatively slowly reacting attenuation elements can be used. The use of the present apparatus in RBV-based systems is particularly advantageous since the decrease in the tip brightness in a significant area of the RBV input screen has a beneficial effect on the contrast in the output image. By the additional reduction of scatter radiation achieved with the use of the inventive apparatus, a lower-noise image results. If necessary, the buffered data of the radiation attenuation, meaning duration, location and degree of the attenuation, can be supplied to a digital image processor that normalizes the contrast over the entire image, as described above. In this application, the attenuation elements operate in intermediate settings (not just in/out) that can be optimally selected based on the information of the preceding images. Multi-slice CT systems represent a further application field of the inventive apparatus. In contrast to conventional CT systems with a single-line detector for image acquisition, the current development trend is in the direction of laminar CT systems. Up to 256 CT slices are simultaneously acquired in such systems by laminar, two-dimensional x-ray detector arrays. The inventive apparatus likewise can be used in such a two-dimensional radiation field as already explained in connection with fluoroscopy applications. In CT systems, however, the absorption data continually change due to the continuous rotation of x-ray detector and focal spot of the x-ray tube. This change can be predicted to a certain extent from the data of the preceding images of the sinogram, such that the respective position of the attenuation elements can be established with suitable prediction electronics. In the simplest version, such prediction electronics assume that the registered translations of the image signals continue further in the preceding images in the sinogram. It is thereby possible to use the same strategies in the control of the attenuation elements as these have already been explained for dose reduction, dynamic increase and/or improvement of the image quality in connection with fluoroscopy. However, since no preceding image data exists at the start of the application, in this case a start condition can be attained, for example with a single pre-scan with reduced dose. By the use of the inventive apparatus in such CT devices, a significant dose reduction as well as an improved image quality result due to reduced scatter radiation intensity. FIG. 1 schematically shows a significantly enlarged section of an embodiment of the present apparatus disposed in the two-dimensional x-ray field of an x-ray image system. The apparatus 1 has a number of attenuation elements 2, arranged in a matrix or grid, that are connected to a carrier substrate 4 via piezoelectric drive elements 3. In this example, the carrier substrate 4 has a number of passage channels 8, with walls to which the drive elements 3 (fashioned as piezo-stack actuators in this example) are attached. Both the webs 9 forming the walls of the carrier 4 through which the passage channels 8 are established as cells and the attenuation elements 2 themselves are, in the rest position, aligned to the focal spot 10 of the x-ray tube, as is to the situation of FIG. 1. The surface of the carrier 4 alternatively can be fashioned in the form of a spherical surface instead of in the shown planar form, such that the x-rays originating from the focal spot 10 are incident perpendicularly at every location on this surface. The attenuation elements 2 are executed as flat paddles or plates of tungsten and each can be tilted around a virtual rotation axis 11 by means of the respective piezo-actuators 3. The actuators 3 responsible for each individual element 2 are arranged within the channel 8 such that they operate in the same direction. They simultaneously expand or simultaneously contract when a corresponding voltage is applied. Since the pair of actuators 3 for element 2 are offset with respect to the virtual rotation axis, the simultaneous expansion or contraction of the pair of actuators 3 tilts or pivots the element 2 as indicated by the curved arrows. The deflected position of the element 2 is shown dashed in the center cell of FIG. 1. In this tilted position, the maximum possible attenuation of x-ray radiation is achieved in the cell. By activation of the piezo-actuators 3 with a lower voltage, arbitrary intermediate positions can also be realized. The current positions of the respective piezo-actuators 3 and thus of the respective attenuation elements 2 is detected in this example by tensiometer strips 6 applied on the piezo-actuators 3. In the neutral position of the attenuation elements, as illustrated with the solid lines, the maximum possible portion of the x-ray radiation is allowed through the present apparatus. The material of the carrier 4 can be selected such that it absorbs x-ray radiation either very significantly or very weakly. In the first case, a fixed ratio of attenuation of the x-ray radiation must always be accepted, while in the second case the x-ray radiation through the apparatus cannot be completely blocked in the shown embodiment. The elements 2 preferably are slanted at their end surfaces such that they lie flat against the walls of the webs 9, as shown by the dashed-line pivoted element 2 in FIG. 1. The x-ray radiation is optimally attenuated by this embodiment given a completely deflected element 2 in the passage channel 8. Since the apparatus is designed for the activation of the drive elements 3 to operate in the same direction, the walls of the webs 9 can serve as electrical terminals (poles) for applying the voltage. The elements 2 thus do not have to be provided with electrical contacts. The contacting of the piezo-actuators 3 can be realized easily in this example, using thin metallic conductor runs on the webs 9 directly toward the edge of the apparatus in parallel or in a number of layers on one side of the carrier 4 while a common electrode is fashioned on the opposite surface of the carrier 4. The surface of the carrier 4 directed toward the side of the focal spot preferably carries the common electrode while the side of the carrier 4 facing away from the focal spot 10 carries the individual conductor runs, since a greater conductor run cross-section can be achieved on this side. In this exemplary embodiment, multi-layer ceramics are used as the piezoelectric actuators 3 because these generate many times the excursion of single-layer ceramics. In order to reduce the requirements on the ceramic excursion in the present embodiment, the actuators 3 should act optimally close to the rotation axis 11, such that a small excursion of the actuators 3 effects a large displacement of the element 1 by lever action. The present apparatus has a number of attenuation elements 2, arranged in a grid, that the respective passage channels 8 of the carrier 4. In this manner, a matrix of controllable absorption cells 12 is formed as can be seen in section in the plan view in FIG. 2. FIG. 2 shows the webs of the carrier 4 that border the passage channels 8. Paddle-shaped attenuation elements 2 that are connected with the walls of the carrier 4 via the piezo-actuators 3 can be seen within the passage channels 8. The attenuation elements 2 are, in this example, held only the actuator elements 3. Such an apparatus can be realized with any desired number of absorption cells 12. For example, a matrix can have 10×10 or 100×100 such absorption cells 12. Since a certain wall thickness of the webs 9 of the carrier 4 is necessary for the stability of the apparatus, it can be advantageous to arrange two or more such apparatuses in succession in the beam direction. A finer degree of spatial modulation of the beam profile is achieved by the multiple attenuation planes obtained in this manner. A particularly advantageous arrangement is achieved when the channels 8 of the two planes disposed in sequence influence are equally large quadratic solid angle of the focal spot 10 of the x-ray tube and are arranged such that one plane influences the light fields of a (theoretical) checkerboard pattern and the other plane influences the dark fields. In an embodiment, the matrix or grid of the absorption cells 12 is disposed within the x-ray image system such that it faces the image matrix of the x-ray detector. The individual attenuation elements 2 of the present apparatus are electronically controlled dependent on the image content of the regulating x-ray image so that a leveling of homogenization of the contrast in the x-ray image is effected. In lighter image regions, the beam attenuators 2 are set to effect a reduced quantum flow, meaning stronger attenuation, while the neutral setting is maintained in darker image regions. The contrast reduction thereby effected at the x-ray detector must be electronically compensated for the image reproduction. For this purpose, the angle setting of the attenuation elements 2 is detected in real time with the sensors 6 and the attenuation linked with the angle setting is used for normalization of the x-ray image. An embodiment of the present apparatus is shown in FIGS. 3 and 4. Only three attenuation elements 2 are shown in side view in FIGS. 3 and 4, but naturally a larger number are present in the actual apparatus. FIG. 3 shows the arrangement of attenuation elements 2 on a (in this example) flat carrier substrate 4 made of a material transparent for x-rays, for example plastic. The carrier 4 alternatively can be spherical, such that the rays from the focal spot 10 of the x-ray tube always strike perpendicularly on the carrier surface. In the present embodiment, the piezoelectric drive elements 3 are executed as flex transducers 5 that are aligned in the direction of the focal spot 10 and stand on the carrier substrate 4. Each flex transducer 5 is preferably tongue-shaped or rod-shaped. The flat attenuation elements 2 that, in this example, are formed of tungsten and preferably exhibit a paddle or plate shape, are attached to the free ends of these flex transducers 5. The connection between the flex transducers 5 and the self-supporting attenuation elements 2 can be realized by gluing, pressing or soldering and only ensues in an end region of the flex transducer 5, which is indicated in FIGS. 3 and 4 with the reference character F. Each attenuation element 2 also forms an absorption channel of the apparatus together with the neighboring element 2. Electrical contacting of the flex transducer 5 ensues on one or both surfaces of the carrier substrate 4, similar to that explained in connection with FIGS. 1, 2 and 3. FIG. 3 shows the neutral setting of the attenuation elements 2 in which these are aligned to the focus 10 of the x-ray tube. A corresponding sensor to detect the curvature (flexing) is mounted on each flex transducer 5, as explained in further detail using FIGS. 7 and 8. The sensors 6, 7 are not shown in FIGS. 3 and 4 (nor in FIGS. 5 and 6, in which they are also used. Given an activation of the piezoelectric flex transducers 5, the attenuation elements 2 are tilted into the beam path of the x-ray radiation, as can be seen in FIG. 4. In this state, the entirety of the radiation is absorbed by the attenuation elements 2. The matrix-like arrangement of these attenuation elements 2 ensues in the same manner as explained in connection with FIGS. 1 and 2. In the embodiment of FIGS. 3 and 4, however, no passage channels are necessary in the carrier substrate 4 since the piezoelectric flex transducers 5 are arranged (with the attenuation elements 2 connected with them) directly on the surface of the substrate 4. FIGS. 5 and 6 show an embodiment of the inventive apparatus comparable to FIGS. 3 and 4, wherein in the attenuation elements 2 are directly fashioned as flex transducers 5. The flex transducers 5 can either be formed directly from a material that strongly absorbs x-ray radiation, for example lead zirconate titanate (PZT), lead metaniobate (PN) or lead nickel niobate (PNN), or can be coated with a layer of such a material, for example tungsten. Otherwise the same features as described in connection with FIGS. 3 and 4 are valid for FIGS. 5 and 6. FIG. 7 shows an example for a flex transducer 5 that can either be used as a piezoelectric drive element 3 for a self-supporting attenuation element 2 according to FIGS. 3 and 4, or directly as an attenuation element 2 according to FIGS. 5 and 6. At a region of the flex transducer 5 that is piezoelectrically influenced, meaning it can be mechanically varied by the application of an electrical voltage, a tensiometer strip 6 is attached with which the curvature of this flex transducer 5 can be detected by expansion or compression of the corresponding region. The tensiometer strip 6 can either be glued on or imprinted. The tensiometer strips 6 connected to measurement electronics 20 with which the deflection of the flex transducer 5 can be quantitatively determined. FIGS. 8A and 8B show a further example of such a flex transducer 5. In this embodiment, a double-layer flex transducer 5 is used, known as a trimorph flex transducer. With this transducer type, curvature is effected by application of an electrical voltage (via the controller 19) on the first layer of the flex transducer (the actuator). The second layer 7 of the flex transducer 5 serves as a sensor that emits a signal with which the curvature is quantitatively determined by the measurement electronics 20. FIG. 8A shows the basic embodiment of such a flex transducer 5 that, as in FIG. 7 as well, can be used either directly as an attenuation element 2 or as a piezoelectric drive element 3 for a self-supporting attenuation element 2. FIG. 8B shows (significantly schematized) the cross-section of such a flex transducer 5 with the additionally integrated piezoelectric layer 7 for the detection of the curvature. A coating 24 of a material that strongly absorbs x-ray radiation is indicated dashed. This coating is provided in the event that the flex transducer 5 is directly used as an attenuation element 2 and is not itself formed of material strongly absorbing x-ray radiation. Furthermore, FIG. 8A shows an embodiment in which the curvature measured by the sensor and quantitatively determined by the measurement electronics 20 used in order to deflect each attenuation element to a desired degree in the form of a control loop. For this purpose, the measurement electronics 20 are connected to the controller 19 to form a control loop. The apparatus described in the exemplary embodiments can be advantageously produced with techniques based on stereolithography. No tools or molds are necessary since changes as well as the design of these apparatuses can be realized on the software level. The carrier substrate in this case is formed of a polymer material, namely a suitably radiation-resistant polymer in order to achieve an acceptable lifespan of the apparatus. A further advantage of the technique of stereolithography for the production of the present apparatus is that the webs of the embodiment according to FIGS. 1 and 2 can be formed such that they are reinforced only where necessary for stability. The unwanted base absorption of the apparatus as well as undesirable radiation hardening due to the plastic body are thereby kept as low as possible. FIG. 9 shows as an example an x-ray imaging system in which the inventive apparatus is used. In this system, the control of the attenuation elements 2 of the inventive apparatus 1 ensues according to the intensity distribution in the subject (the patient 16) determined in the detector output signal. FIG. 9 shows the high-voltage generator 13 for the operation of the x-ray tube 14. The patient 16 who is irradiated by the x-rays is positioned between the x-ray tube 14 and the x-ray image detector 17. A typical radiation diaphragm 15 to limit the radiation field as well as the inventive modulation apparatus 1 are disposed on the side near the x-ray tube. The intensity distribution within the image received by the detector 17 is evaluated by detector electronics 18. Given detection of lighter image locations, the attenuation elements are cell-selectively or channel-selectively activated by the controller 19 in order to reduce the dose in particular radiation channels. The position of the individual attenuation elements 22 within the apparatus 1 is detected and processed in real time with the measurement device 20 that is connected to the sensors 6, 7 of the attenuation elements 2, in order to provide the channel-dependent attenuation to a digital image post-processing 22 via a storage unit 21. The real value of the current attenuator setting is stored as a time curve in the storage unit 21. In this manner, the applied dose can be calculated for all pixels. The value for the exact reproduction (normalization) of the contrast values for the image representation of the x-ray image on the screen 23 can be derived from this information, the image representation being executed by digital image post-processing electronics 22. The image signal of pixels that (as a consequence of the setting of the attenuation elements 2) have received less quanta compared to others for which the attenuation elements are completely open (i.e. in the neutral position) is intensified corresponding to the calculated reduction of the quantum flow, thus increased in terms of contrast. The desired homogenous image impression results in this manner. The detected real values of the positions of the attenuation elements 2 can be simultaneously supplied to the attenuator controller 19 in order to form a control loop with which the position of the attenuation elements 2 can be exactly adjusted. Although modifications and changes may be suggested by those skilled in the art, it is the intention of the inventors to embody within the patent warranted hereon all changes and modifications as reasonably and properly come within the scope of their contribution to the art.
	claims	1. A neutron-optical component array for the specific spectral shaping of neutron beams or pulses in a neutron guide or in a radiation hole between a fast neutron source with a plurality of moderators of different structures arrayed closely together for generating slow neutrons of different energy spectra as well as for their radiation in predetermined radiation directions and at least one place of experiment,characterized by the fact thatthe radiation directions (CBL, TBL) of the moderators (CNM, TNM) are overlapped directly or by further neutron-optical components (RSM, SSM) in the neutron guide (NGT) or at the place of experiment and that the slow neutrons (CCN, TTN) generated by the moderators (CNM, TNM) of different energy spectra are integrated in common in an overlapping neutron beam (SBL) with a multi spectrum defined by the structure and number of the moderators (CNM, TNM). 2. The neutron-optical component array according to claim 1,characterized by the fact thatin case of a direct overlapping of the radiation directions they are combinable by a predetermined encoding scheme at the place of experiment. 3. The neutron-optical component array according to claim 1,characterized by the fact thatthe neutron guide (NGT) is coated with nickel on its internal surface (INS). 4. The neutron-optical component array according to claim 1,characterized by the fact thatin case of overlapping of the radiation directions by further neutron-optical components for obtaining an effective mean radiation direction of the overlapping neutron beam a further neutron-optical component is structured as an oscillating reflector which oscillated in synchronism with the pulsed neutron source or with the chopped neutron beam of a continuous neutron source. 5. The neutron-optical component array according to claim 1,characterized by the fact thatin case of overlapping of the radiation directions (CBL, TBL) by further neutron-optical components (NOC) for obtaining an effective mean radiation direction (EBL) of the overlapping neutron beam (SBL) a further neutron-optical component (SSM) is provided with an energy dependent switching function. 6. The neutron-optical component array according to claim 5,characterized by the fact thatthe further neutron-optical component (NOC) is structured as a neutron reflector (SSM) with an energy dependent switching function which by a corresponding angular alignment continuously or intermittently passes or reflects impinging neutrons as a function of their energy. 7. The neutron-optical component array according to claim 5,characterized by the fact thatthe neutron reflectors (RSM, SSM) are structured in a self supporting form or in a form coated on a neutron transparent substrate as a single or multi-layered neutron reflector, with the coating being applied to one of both sides of the substrate. 8. The neutron-optical component array according to claim 4,characterized by the fact thatthe further neutron-optical components (NOC, RSM, SSM) are integrated into the neutron guide (NGT).
	description	This application is a divisional patent application of application Ser. No. 13/451,050, filed Apr. 19, 2012, which Application is a continuation patent application of U.S. patent application Ser. No. 10/966,907, filed Oct. 15, 2004, entitled “ADVANCED FIRST CORE FUEL ASSEMBLY CONFIGURATION AND METHOD OF IMPLEMENTING THE SAME.” Field of the Invention The present invention relates generally to fuel assemblies for a nuclear reactor and, more particularly, to an advanced method of fuel management through arrangement of nuclear fuel assemblies within the initial core of a pressurized water reactor. The invention also relates to advanced initial cores for pressurized water nuclear reactors. Background Information Modern commercial nuclear power reactors are fueled with uranium having a slightly enriched U-235 content. The core of the reactor is formed by numerous elongated, rectangular fuel assemblies arranged in a cylindrical vessel. The fuel assemblies are arranged in accordance with a loading pattern intended to meet certain engineering requirements, such as the distribution of power, including limits on power peaks within the core. Other considerations include the maximization of the fuel cycle, or the time required between refuelings. The initial loading configuration and plan of replacement and arrangement of fuel during the life of the reactor is known as in-core fuel management, and is a major nuclear reactor design consideration. Use of the slightly enriched U-235 fuel necessitates that portions of the core be periodically removed and replaced with new or fresh fuel. Thus, it is common to combine fuel assemblies from previous fuel cycles with new fuel. A typical inventory of fuel assemblies includes about one-third new fuel assemblies, commonly referred to as feed assemblies, about one-third once-burned fuel assemblies and about one-third twice-burned fuel assemblies. Accordingly, the fuel loading pattern for the first core of a nuclear reactor, such as a pressurized water reactor, commonly utilizes three enrichment zones, based upon the average enrichment of U-235 of the fuel assemblies with a given zone, with each zone having generally equal proportions. FIG. 1 shows a schematic representation of such a known prior art loading pattern 2 which uses three generally equal-sized fuel batches 4,6,8. A batch is a group of fuel assemblies that are typically placed into, and then permanently removed from, the core 14 together. Note that the pattern shown in FIG. 1 illustrates only one-eighth of the reactor core 14 and assumes core symmetry. The zones include two low enrichment zones 4, 6 which, as shown, are generally loaded in a checkerboard fashion toward the reactor interior 10, and a high enrichment zone 8, which is loaded primarily at the reactor periphery 12. The length of the fuel reload cycle for cores 14 assembled in accordance with the foregoing arrangement is adjusted by varying the enrichments of all of the zones 4, 6, 8 in a substantially equal manner. However, this results in the disadvantage of relatively poor fuel utilization due primarily to high neutron leakage caused by the high enrichment zone 8 being disposed at the reactor periphery 12. Additionally, the use of three generally equal-sized regions in zones 4, 6, 8 is inconsistent with current industry fuel management practices with regard to reload cycles, wherein the refueling fraction varies as a function of the desired cycle length. A cycle is the time during which the arrangement of normally stationary fuel in the reactor core is unchanged, usually beginning with the placement of a feed batch, or a batch of fresh fuel, into the core, and ending with the removal of highly burned fuel assemblies. The number of burns a fuel assembly has experienced is the number of cycles it has been in the reactor core. A typical cycle might range from 10 to 18 months in duration. By way of example, 18-month cycles in accordance with the aforementioned prior art require about 40% of the core to be replaced at each cycle, with the replacement typically comprising a mixture of fuel assemblies having both low and high initial enrichments. Therefore, it will be appreciated that the use of equal size batches results in the discharge of low enrichment regions or zones at very low burn-up, while requiring a significant financial investment. Accordingly, such practice is inefficient and uneconomical. The greatest savings in overall fuel costs is achieved by minimizing the initial enrichment required to achieve an equilibrium fuel management scheme. Further adding to the inefficiency of known prior art methods for establishing initial cores is the fact that such methods are essentially ad hoc basis, performed on a substantially trial and error basis relying on years of experience in the art. There is no systematic approach for developing the first core. As such, initial cores created by such methods must be conservatively designed, with a built in margin for error, which results in the core taking longer than necessary to reach equilibrium, thereby raising costs. It is desirable, therefore, to avoid the substantial fuel cycle cost penalties associated with known initial core nuclear fuel management schemes. There is a need, therefore, for an advanced method of implementing initial cores for nuclear reactors. Accordingly, there is room for improvement in the art of in-core fuel management for nuclear reactors including initial core fuel assembly arrangement and in methods of implementing the same. These needs and others are satisfied by the present invention, which is directed to a method of loading an initial core for a nuclear reactor such that it substantially emulates the highly economical and efficient equilibrium cycle reload core of a nuclear reactor. As one aspect of the invention, a method of implementing an economical initial core for a nuclear reactor comprises: providing a plurality of fuel assemblies having different average enrichments of uranium 235; and arranging the plurality of fuel assemblies in an initial core configuration structured to emulate a known equilibrium reload cycle core at least in terms of spatial reactivity distribution. The method may include providing bundles of the plurality of fuel assemblies, wherein the bundles include at least highly enriched fuel batches, less enriched fuel batches and least enriched fuel batches. The equilibrium cycle reload core may include batches of feed fuel assemblies, once-burned fuel assemblies and twice-burned fuel assemblies, wherein the feed fuel assemblies have the highest average enrichment of uranium 235, the once-burned fuel assemblies have less enriched fuel and the twice-burned fuel assemblies have the least enriched fuel, and wherein the highly enriched fuel batches of the initial core are approximately the same size and enrichment as the feed fuel batch of the equilibrium cycle reload core. The method may further include the step of approximating the reactivity of the once-burned fuel batches and the twice-burned fuel batches of the equilibrium cycle reload core in the initial core through use of fuel batches in the initial core which have initial average enrichments based upon the reactivity of the non-zero, beginning of cycle burnup and initial enrichment of the once and twice-burned fuel of the equilibrium reload cycle core. One or more of the fuel batches of the initial core may include one or more sub-batches comprising fuel assemblies of similar average enrichment. The method may include providing as the highly enriched fuel batches, two high enrichment sub-batches sized and enriched to emulate the feed fuel batches of the known equilibrium reload cycle core. The high enrichment sub-batches may be loaded toward the interior of the initial core and the low enrichment batches of fuel may be primarily loaded at the periphery of the initial core. The method may include providing an advanced lattice design for the initial core wherein the lattice design is structured to compensate for a large thermal neutron flux peak which results from the highly enriched fuel batches within the initial core configuration being disposed adjacent less enriched fuel batches within the initial core configuration. The advanced lattice design may include a generally square pattern of adjacent rows of fuel rods wherein the fuel rods include at least six different fuel rod types ranging in average enrichment from least enriched to most enriched and wherein the advanced lattice disposes the least enriched fuel rods at the corners of the generally square pattern, the next to least enriched fuel rods at the outermost rows of the generally square pattern and the most enriched fuel rods toward the center of the generally square pattern. As another aspect of the invention, an advanced initial core for a nuclear reactor comprises: a plurality of fuel assemblies having different average enrichments of uranium 235, wherein the initial core is loaded by arranging the plurality of fuel assemblies in a configuration that emulates a known equilibrium reload cycle core, at least in terms of spatial reactivity distribution. The advanced initial core may include bundles of the plurality of fuel assemblies wherein the bundles include at least highly enriched fuel batches, less enriched fuel batches and least enriched fuel batches. The equilibrium cycle reload core may include batches of feed fuel assemblies, once-burned fuel assemblies and twice-burned fuel assemblies wherein the feed fuel assemblies have the highest average enrichment of uranium 235, the once-burned fuel assemblies have less enriched fuel and the twice-burned fuel assemblies have the least enriched fuel and wherein the highly enriched fuel batches of the initial core are approximately the same size and enrichment as the feed fuel batch of the equilibrium cycle reload core. The reactivity of the once-burned fuel batches and the twice-burned fuel batches of the equilibrium cycle reload core may be approximated in the initial core through use of fuel batches in the initial core which have initial average enrichments based upon the reactivity of the non-zero, beginning of cycle burnup and initial enrichment of the once and twice-burned fuel of the equilibrium reload cycle core. One or more of the fuel batches of the initial core may include one or more sub-batches comprising fuel assemblies of similar average enrichment. The highly enriched fuel batches of the initial core may include two high enrichment sub-batches sized and enriched to emulate the feed fuel batches of the known equilibrium reload cycle core. The high enrichment sub-batches may be loaded toward the interior of the initial core and the low enrichment batches of fuel may be primarily loaded at the periphery of the initial core. The advanced initial core may include an advanced lattice design wherein the lattice design is structured to compensate for a large thermal neutron flux peak which results from the highly enriched fuel batches within the initial core configuration being disposed adjacent less enriched fuel batches within the initial core configuration. The invention will be described as applied to a method of implementing initial cores for pressurized water nuclear reactors using data for an 18-month equilibrium cycle reload core as a design target, although it will become apparent that it could be equally applicable to other fuel reloading cycles regardless of the number of fuel batches used. It will also be appreciated that the figures and corresponding disclosure herein merely provide representative examples of initial cores generated through application of the method of the present invention and are, therefore, not limiting upon the scope of the invention. Before proceeding with a more detailed description of the present invention, it is important to provide an overview of terminology commonly employed in the nuclear power industry, as it will be applied herein. Certain such terms were described in the Background Information section hereinbefore. Additional terminology will now be discussed. As employed herein, the reference to “checkerboard” is used to describe a known prior art reactor core fuel assembly loading pattern wherein fuel assemblies of like enrichment are arranged in a manner analogous similar to the grid of red and black squares on a checkerboard, with fuel assemblies of one enrichment (e.g., once-burned fuel) being disposed in a position corresponding to the red squares of the checkerboard, and fuel assemblies of another enrichment (e.g., twice-burned fuel) being disposed in the staggered, adjacent positions corresponding to the black squares of the checkerboard. The term “equilibrium” is typically used with reference to the highly desirable, substantially stable state of the reactor core which is traditionally achieved after several fuel reload cycles following the initial core cycle. Once achieved, the equilibrium cycle provides very efficient operation (e.g., fuel burn) and predictable, economical reloading of the core at subsequent reload cycles. Accordingly, it is usually desirable to achieve an equilibrium in-core fuel management scheme as early as possible in the plant lifetime. In equilibrium, the feed assemblies will always have substantially the same enrichment and will be placed in the same locations as the previous feed assemblies and the once-burned and twice-burned assemblies that remain in the core, will be shuffled to identical locations occupied by the previously once and twice-burned assemblies. As employed herein, the term “bundle” refers to one or more than one (i.e., a plurality) fuel assembly. The term “lattice” refers to the array of fuel within a fuel assembly and the particular arrangement of individual fuel rods within such array. The advanced fuel assembly configuration of the present invention results in a first core loading pattern that closely emulates the spatial reactivity distribution of such desirable equilibrium reload core. Comparing the initial core of the present invention with, for example, the known prior art initial core fuel assembly arrangement of FIG. 1, and with the desired equilibrium core of FIG. 2, which is typically achieved after several reload cycles following the initial core cycle of FIG. 1, will help to provide a clear understanding and appreciation of the present invention. FIG. 1, as previously discussed, shows a known prior art reactor core loading pattern 2, as represented in a schematic diagram of one-eighth of the reactor core 14 which is assumed to be symmetric in shape. FIG. 1A shows the outline of the entire core 14 and the location of the one-eighth section thereof, shown in FIG. 1. As shown in FIG. 1, the loading pattern 2 places the highest enrichment zone 8, which includes fuel represented symbolically with the letter A, at the periphery 12 of the core 14. Zones 4, 6 include less reactive, once-burned fuel B and least reactive, twice-burned fuel C, respectively, which are disposed in a checkerboard pattern toward the interior 10 of the core 14, as shown. Such configuration, as previously discussed, results in a reactor power distribution that is peaked near the core periphery 12, where neutrons will leak out of the reactor and never return. Accordingly, the aforementioned fuel assembly arrangement results in poor fuel utilization and economical inefficiency. The advanced initial core (e.g., 214 of FIG. 3) and method of implementing the same, of the present invention, overcome these shortcomings by providing an initial core 214 (FIG. 3) that permits a lower initial enrichment while providing the same energy extraction and establishes a power distribution which remains centrally peaked throughout the burnup cycle. Specifically, the present invention uses the desirable equilibrium cycle reload core 114 (FIG. 2) as a design target and emulates at least the reactivity distribution of the equilibrium core 114 in the initial reactor core 214 (FIG. 3). The reactivity distribution of the equilibrium core 114 (FIG. 2) provides a model for determining the number of discrete regions within the first core loading pattern 216 (FIG. 3) that emulate the fresh (e.g., fuel assemblies A′ of zone 108 in FIG. 2), once-burned (e.g., fuel assemblies B′ of zone 104 in FIG. 2) and twice-burned fuel (e.g., fuel assemblies C′ of zone 106 in FIG. 2) of the target equilibrium reload cycle core 114. The reactivity values for the equilibrium core 114 are obtained through the initial enrichment of the fuel assemblies A, B, C and, to a lesser extent, from the burnable absorber loading of such assemblies. This can be more fully understood by comparing FIGS. 2 and 3, which will now be discussed in further detail. FIG. 2 shows a schematic representation of the loading pattern 116 of the target equilibrium cycle reload core. Like FIG. 1, FIG. 2 also assumes one-eighth core symmetry. As shown, unlike the high enrichment fuel A of zone B in the prior art loading pattern 2 of FIG. 1, the high enrichment fuel assemblies A′ are located toward the core interior 110, in zone 108. The highly enriched fuel A′ is checkerboarded with once-burned fuel assemblies B′, also disposed towards the core interior 110, in zone 104. The least enriched fuel assembles C′ are primarily disposed on the core periphery 112. FIG. 3 shows a schematic illustration of a representative first core loading pattern 216 and resulting initial core 214 when employing the advanced method of the present invention in order to emulate the desired equilibrium cycle reload core 114 of FIG. 2. Specifically, in accordance with the advanced method of the present invention, the reactivity distribution of the desirable equilibrium cycle reload core 114 is known. Such distribution is illustrated in the example of FIG. 2. Additionally, it is well known that much data, including data relating to such reactivity distribution, is available with respect to equilibrium cycles for nuclear reactors, generally. Such data has been obtained through years of monitoring the equilibrium cycles of various reactors and through the development and application of many known prior art methods (e.g., algorithms) in the nuclear art over the years in an attempt to optimize in-core fuel management. In essence, the advanced method of the present invention involves taking such data and applying it to emulate the reactivity distribution of the highly economic and efficient equilibrium cycle reload core 114 (FIG. 2), in the initial core 214. In other words, the method of the present invention emulates a desired end result (e.g., core equilibrium) which typically takes years to achieve, and achieves a comparable result in an initial step (e.g., the initial core 214 of the reactor), thereby dramatically improving fuel management efficiency and economy. This involves applying known parameters of the desired equilibrium cycle reload core 114, such as the initial enrichment of particular fuel assemblies 104, 106, 108 and the burnable absorber loading or amount of burnup of the assembly, and emulating the reactivity distribution which results from such parameters, using only those fuel assembly parameters which may be varied within the initial core 214. For example, only initial enrichment, batch quantity and location of the fuel assemblies within the core, may be varied because all other parameters are at an initial start up value of zero variability, the initial core has not yet experienced any burnup. Therefore, the method of the present invention backs out an advanced core 214 using known parameters from the desired equilibrium target core and replicating them as closely as possible. FIG. 3, again illustrates one-eighth of the core 214 and assumes core symmetry. Comparing the initial core 214 of FIG. 3 with the equilibrium cycle reload core 114 of FIG. 2, it is evident that the exemplary initial core 214 succeeds in closely emulating the spatial reactivity distribution of the very economical target equilibrium cycle reload core 114 both in terms of sub-batch quantity and spatial distribution. In order to accomplish this goal, the method of the present invention, unlike the three substantially equal zones of known initial cores (e.g., zones 4, 6 and 8 of FIG. 1), provides a core having essentially any necessary, suitable combination and configuration of fuel batches and sub-batches. Specifically, the enrichments of the exemplary initial core 214 are obtained through use of a group of high enrichment bundles 208 of approximately the same size and enrichment as the equilibrium cycle feed batch 108. Second, the burned fuel 104, 106 in the equilibrium reload cycle core 114 is approximated through use of initial enrichments that approximate the reactivity of the non-zero beginning of cycle (BOC) burnup and initial enrichment of the burned fuel inventory 104, 106 of the equilibrium reload cycle core 114. The resulting enrichment range of the initial core 214 is much larger than the known prior art (see, e.g., FIG. 1). For example, the enrichment of the exemplary initial core 214 of FIG. 3 ranges from about 0.30 to above about 4.5, as measured in weight of U-235. Conversely, by way of comparative example, the representative prior art initial core 14 of FIG. 1 has an enrichment range from about 2.35 to about 4.45 weight of U-235. Furthermore, the exemplary initial core 214 of FIG. 3 includes two high enrichment sub-batches 208, 208′ which are sized and enriched to emulate the fresh fuel region or zone 108 of the equilibrium reload cycle 116. The next two sub-batches 204, 204′ are sized similarly to and are enriched to emulate the reactivity of the once-burned portion 104 of the equilibrium cycle reload core 114. Finally, the remaining sub-batches, for example, 206, 206′ are sized to reflect the twice-burned portion 106 of the reload core 114 with enrichments being similarly selected to emulate the reactivity of the twice-burned fuel 106 in the equilibrium reload core 114. In the example of FIG. 3, the two sub-batches of highly enriched fuel 208, 208′ are positioned toward the interior 210 of the core 214 rather than at the periphery 212 like the known prior art (see, e.g., highly enriched fuel 8 at core periphery 12 of FIG. 1). The method next involves positioning the pair of sub-batches of fuel 206, 206′ emulating the twice-burned fuel 106 of the equilibrium cycle reload core 114 primarily at the periphery 212 of the core 214. The remainder of the fuel within the initial core 214 can be enriched and arranged as needed, in any combination, in order to replicate the desired equilibrium cycle. For example, the initial core 214 shown in FIG. 3 includes two sub-batches 204, 204′ sized and enriched to emulate the once-burned fuel 104 of the equilibrium core 214. Such sub-batches 204, 204′ are disposed intermittently, as necessary to achieve the desired reactivity distribution. Accordingly, the initial core 214 and method of implementing the same, of the invention, overcomes the disadvantageous neutron leakage of the prior art, premature removal or replacement of highly enriched fuel and the economic inefficiencies associated therewith. Additionally, because the initial core 214 emulates the reactivity distributions of the equilibrium reload core 114, the time or cycle length to transition to the reload core cycle is dramatically improved. Such improvement is due largely to the much wider range of enrichments and to the unequal sub-batch sizes used in the advanced initial core 214. This set of advanced core design parameters has the effect of minimizing the discharge enrichment of the initial core 214 and, thereby, maximizing the enrichment of the fuel carried over for further exposure within the reactor. The fuel which is carried over from the initial cycle 214 to the reload cycles is of an enrichment and reactivity that is representative of the target reload core 114. In order to support the foregoing advanced initial core design 214, an advanced lattice design is required. The lattice is the array of fuel and the particular arrangement of rod locations in a fuel assembly. FIG. 4 shows a representative example of such an advanced lattice design 300 for at least some of the fuel assemblies of the exemplary initial core 214. For ease of illustration, only the upper left quadrant of the lattice 300 is shown. Mirror symmetry in the remaining quadrants is assumed (see FIG. 4A, highlighting the location of the one-quarter section of FIG. 4 within the lattice 300). The lattice 300 is designed to compensate for the highly enriched inboard fuel bundles 208, 208′ that are adjacent to low enrichment bundles (e.g., 206, 206′). It is well known that such a fuel loading arrangement will result in a large thermal neutron flux peak at the radial edge rows 304 of the high enrichment bundle 208 due to the very low absorption of the low enrichment bundles 206 (not shown in FIG. 4). This is one aspect of the equilibrium reload core 114 (FIG. 2) that cannot be emulated in the initial core 214 (FIG. 3) because the equilibrium reload core low-reactivity bundles (e.g., 106) are much more highly absorbing of neutrons than the initial core low-enrichment bundles (e.g., 204). This is because the burned bundles of the reload core 114 include highly absorbing fission products and high-order actinides. The initial core low enrichment bundles (e.g., 206, 206′) having not been burned, do not possess such absorbing nuclear products. In known prior art lattice designs (not shown), each fuel rod in a given assembly generally has the same enrichment. A uniformly enriched bundle design, for the highly enriched bundles, would result in unacceptably high power peaking factors within the exemplary initial core 214 (FIG. 3). Thus, the advanced lattice design 300 of the present invention uses six different rod types (see, e.g., the key beneath lattice 300 of FIG. 4) in a configuration designed to reduce peaking factors when the highly enriched assemblies are adjacent lower enriched assemblies (see, e.g., fuel assemblies 208 adjacent fuel assemblies 206′ of FIG. 2). Lower enriched in this case refers to enrichments ranging from between about 1 to 2.50 weight of U-235, in order to simulate the reactivity of burned assemblies (e.g., 104, 106). Specifically, as shown in FIG. 4, radial zoning of the bundle 208 is employed to reduce the enrichment of the fuel rods 302 at the edges of the bundle 208 and, thereby, result in a bundle power distribution that compensates for the unusually high thermal neutron flux at the bundle periphery 212 (FIG. 3). Such enrichment reduction is typically limited to the first two rows 304, 306 of the bundle 208 due to the thermal neutron mean-free path in the high enrichment bundle 208 which is relatively short (e.g., roughly one pitch). The example of FIG. 4 shows the lattice 300 or how the fuel rods are arranged to form a bundle, such as the high enrichment bundle 208, shown. Six types of fuel rods 302, 304, 306, 308, 310, 312 are employed and range, respectively, in level of enrichment from least enriched to most enriched. As previously discussed, low enrichment in the exemplary case refers to between about 0.3-2.50 weight-percent of U-235. Medium enrichment is preferably between about 2.50-3.50 weight of U-235 and high enrichment is preferably between about 3.50-5 weight of U-235. It will, however, be appreciated that there are variations in enrichment even with these designated categories (e.g., low, medium, high). This will be appreciated with reference to the various enrichments of the fuel assemblies (e.g., A,B,C,A′,B′,C′) and bundles (e.g., 4,6,8,104,106,108,204,206,208) in FIGS. 1-3 and of the individual fuel rods (e.g., 302,304,306,308,310,312) as designated in the legend under lattice 300 of FIG. 4, wherein VL designates very low enrichment fuel rods 302, L designates low enrichment fuel rods 304, M designates medium enrichment fuel rods 306, MI designates medium enrichment fuel rods with integral fuel burnable absorbers 308, H designates high enrichment fuel rods 310, and HI designates high enrichment fuel rods with integral fuel burnable absorbers 312. It will further be appreciated that the specific enrichment levels and ranges illustrated and discussed herein are merely an example for an initial core (e.g., 214) loaded in accordance with the present invention. A wide variety of alternative arrangements are within the scope of the invention. The lattice 300 of FIG. 4 has a generally square pattern of adjacent fuel rods. The lowest enrichment fuel rods or very low (VL) enrichment fuel rods 302 are disposed at the corners (one corner, and thus one very low (VL) enrichment fuel rod 302, are shown in FIG. 4) of the lattice 300. The next lowest enrichment rods or low (L) enrichment fuel rods 304 are disposed along the sides or edges of the lattice 300. The fuel rod enrichments then gradually increase as the rod positions approach the center of the lattice 300 where the most highly enriched fuel rods or high enrichment fuel rods with integral fuel burnable absorbers (HI) 312 and high (H) enrichment fuel rods 310 and next to most highly enriched or medium enrichment fuel rods with integral fel burnable absorbers (MI) 308 are disposed in a symmetric pattern. Accordingly, the foregoing advanced lattice design 300 of the present invention compensates for the thermal flux peak of the environment, primarily caused by the inboard highly enriched bundles 208, 208′ (best shown in FIG. 3) of the exemplary initial core 214 (FIG. 3) being adjacent low enrichment bundles 206, 206′ (best shown in FIG. 3), by lowering the enrichment at the edges or sides 304 of the lattice 300, thereby reducing such power peaks and supporting the highly efficient and thus economical initial core 214 (FIG. 3) of the present invention. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular arrangements disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the claims appended and any and all equivalents thereof.
	claims	1. A container for the consolidation of material under elevated pressure and temperature conditions, comprising:an outer cylinder; andan inner cylinder comprising internal compression plates that are positioned between the inner and outer cylinders and curved to match the radius of the inner and outer cylinders, wherein said compression plates are arranged in rows and with predetermined spacing, axially, radially, or both. 2. The container of claim 1, wherein the material comprises hazardous, toxic, or radioactive waste, and the container is configured to hold such waste without releasing it to the environment. 3. The container of claim 1, wherein the inner and outer cylinders are made from metal comprising steel, nickel, titanium, aluminum, copper, alloys thereof, or combinations thereof, wherein the inner cylinder has at least one different characteristic from the outer cylinder, said characteristic comprising malleability, corrosion resistance, or wall thickness. 4. The container of claim 1, wherein the inner cylinder comprises a layer that is chemically reactive with the material located in the container. 5. The container of claim 4, wherein said layer comprises titanium in an amount sufficient to (i) react with oxygen that degases from the waste material being consolidated, (ii) control the redox of the powdered waste material, or (iii) combinations thereof. 6. The container of claim 1, wherein the outer cylinder has a wall thickness that is thicker than the inner cylinder. 7. The container of claim 1, wherein the internal compression plates comprise a material that has a higher strength than the inner cylinder, the outer cylinder, or both, such that it resists collapse and deformation under hot isostatic pressing conditions, wherein said material comprises a metal, ceramic, graphite or combinations thereof. 8. The container of claim 7, wherein the internal compression plates are configured to touch at least one other compression plate during hot isostatic pressing. 9. The container of claim 8, wherein the internal compression plates comprise right angled edges. 10. The container of claim 7, wherein the internal compression plates have angled or recessed edges to cause interlocking or guide the plates to slide over each other during hot isostatic pressing. 11. The container of claim 1, further comprising a liner configured around the compression plates that help lock the plates into position. 12. The container of claim 1, wherein the outer cylinder comprises walls of sufficient thickness to allow the walls to remain straight after being exposed to said elevated pressure and temperature conditions. 13. A method of producing a consolidated article, the method comprising:filling a container with material to be consolidated, the container comprising:an outer cylinder;an inner cylinder comprising internal compression plates that are positioned between the inner and outer cylinders and curved to match the radius of the inner and outer cylinders, wherein said compression plates are; andcollapsing the by applying heat and/or pressure to the container such that the internal compression plates cause the container to collapse in a predictable manner while consolidating the material in the container to produce a consolidated article having a predictable shape and/or dimension. 14. The method of claim 13, further comprising evacuating and sealing the container prior to consolidating. 15. The method of claim 13, wherein the material comprises hazardous, toxic, or radioactive waste, and the container is configured to hold such waste without releasing it to the environment. 16. The method of claim 13, further comprising configuring the plates to resist collapse during consolidation are lined up in rows and with predetermined spacing both axially and radially. 17. The method of claim 13, further comprising reacting the material to be consolidated with at least one material located on or within the inner cylinder. 18. The method of claim 17, wherein the material located on or within the inner cylinder comprises titanium, and said reacting comprises (i) reacting with oxygen that degases from the waste material being consolidated, (ii) controlling the redox of the powdered waste material, or (iii) combinations thereof. 19. The method of claim 13, wherein collapsing the by applying heat and/or pressure to the container comprises hot isostatic pressing at a temperature ranging from 800 to 1400° C. and pressure ranging from 10-300 MPa for a time ranging from 8 to 14 hours.
	abstract	The present invention relates to a method for removing cesium in a clay mineral using hydrogen peroxide. According to the present invention, cesium in a clay mineral is removed using hydrogen peroxide, which serves to induce interlayer expansion of the clay mineral to allow a cation to easily enter an interlayer of the clay mineral, and thus cesium desorption efficiency can be further improved. Also, the method according to the present invention can be efficiently used to restore soil in residential areas widely contaminated with a radionuclide when a major accident such as Fukushima nuclear accident occurs as well as various sites of atomic energy facilities contaminated with a radionuclide. Also, since radiation-contaminated soil is treated with only hydrogen peroxide and cations, secondary environmental pollution caused by wastes can be significantly reduce and the waste disposing cost can also be saved.
	claims	1. A method of acquiring an image of a specimen comprising: (a) creating a charging map of said specimen including (i) first scanning said specimen for applying a first dose of a particle beam and storing corresponding detected first pixel intensities; (ii) second scanning said specimen applying a second dose of a particle beam greater than said first dose and storing corresponding detected second pixel intensities; (iii) subtracting said first and second pixel intensities to form values indicating charging areas and non-charging areas; and (b) selectively scanning said specimen based on said charging map after creating said charging map, to obtain an image of said specimen wherein charging areas are given a reduced dose of a particle beam relative to non-charging areas. 2. A method as recited in claim 1 wherein said first scanning is performed more rapidly than said second scanning for achieving said greater second dose. claim 1 3. A method as recited in claim 1 wherein a beam intensity of said second scanning is greater than a beam intensity of said first scanning for achieving said greater second dose. claim 1 4. A method as recited in claim 1 wherein said selectively scanning includes a first plurality of scans wherein each said charging area determined by said charging map is irradiated by a beam only during a calculated percentage of said first plurality of scans. claim 1 5. The method as recited in claim 1 wherein said non-charging areas are given a maximum dose of the particle beam. claim 1 6. The method as recited in claim 1 wherein said particle beam is an electron beam. claim 1
	description	This Patent Application claims the priority benefit of U.S. provisional patent application No. 62/485,916, titled “TALBOT X-RAY MICROSCOPE,” filed Apr. 15, 2017, and is a continuation-in-part of U.S. patent application Ser. No. 14/712,917, filed May 15, 2015 and entitled “X-RAY METHOD FOR MEASUREMENT, CHARACTERIZATION, AND ANALYSIS OF PERIODIC STRUCTURES”, which in turn is a continuation-in-part of U.S. patent application Ser. No. 14/700,137, filed Apr. 29, 2015 and entitled “X-RAY INTERFEROMETRIC IMAGING SYSTEM”, which in turn is a continuation-in-part of U.S. patent application Ser. No. 14/527,523, filed Oct. 29, 2014 and entitled “X-RAY INTERFEROMETRIC IMAGING SYSTEM”, which in turn claims the benefit of U.S. Provisional Patent Application Nos. 61/898,019, filed Oct. 31, 2013 and entitled “X-ray Phase Contrast imaging System”; 61/901,361, filed on Nov. 7, 2013 and entitled “An X-ray Source Consisting of an Array of Fine Sub-Sources”; and 61/981,098, filed Apr. 17, 2014 and entitled “Two Dimensional Phase Contrast Imaging Apparatus”, the disclosures of all of which are incorporated herein by reference in their entirety. Application Ser. No. 14/712,917 also claims the benefit of 61/993,792 filed May 15, 2014 and entitled “Method of Talbot-Effect Based X-ray Patterned Probe and Characterization (Metrology or Inspection) Apparatuses Using Such”, the disclosure of which is hereby incorporated by reference in its entirety. The present technology relates to interferometric systems using x-rays, and in particular, interferometric measurement, characterization and analysis systems using a system of periodic micro-beams to illuminate an object to determine various structural and chemical properties of the object. Prior art x-ray microscopes are generally limited by the resolution of the x-ray optics (e.g. zone plates) and/or the resolution of the pixel size of the detector. Although some commercial x-ray microscope systems have a resolution of less than 100 nm, such systems have an extremely limited field of view, and high resolution x-ray microscopy with a large field of view has difficulty producing images with a resolution smaller than 1 micron. Talbot systems of the prior art have traditionally been used for low resolution imaging. What is needed is a microscopy system that utilizes Talbot interference fringes for high resolution imaging at improved throughput. This present technology, roughly described, includes systems for x-ray microscopy using an array of micro-beams having a micro- or nano-scale beam intensity profile to provide selective illumination of micro- or nano-scale regions of an object. An array detector is positioned such that each pixel of the detector only detects x-rays corresponding to a single micro-beam, allowing the signal arising from the x-ray detector to be identified with the specific, limited micro- or nano-scale regions illuminated. This enables microscopy while using a higher efficiency, larger pixel detector without compromising spatial resolution. In embodiments, the micro- or nano-scale beams may be provided by producing a set of Talbot interference fringes, which creates a set of fine x-ray micro-beams corresponding to beam comprising the anti-nodes of the interference pattern. In some embodiments, the array of micro- or nano-beams may be provided by a conventional x-ray source and an array of x-ray imaging elements (e.g. x-ray lenses). In embodiments, both the detector and the object are placed within the same waist or “depth-of-focus” range of a set of Talbot constructive fringes (anti-nodes). In some embodiments, the detector is placed downstream at any subsequent set of anti-nodes (an integer number of Talbot distances away). In some embodiments, the object is positioned on a mount that allows translation in the x- and y-directions perpendicular to the direction of x-ray beam propagation, allowing a “scanned” transmission image on a microscopic scale to be assembled. In some embodiments, the object is positioned on a mount that allows rotation about an axis perpendicular to the direction of x-ray beam propagation, allowing the collection of data on a microscopic scale to be used for laminographic or tomographic images reconstruction. Additional masking layers may be inserted in the beam path to block a selected number of the micro-beams, allowing the use of detectors with larger pixel sizes for the remaining micro-beams. The use of a masking layer also allows the use of a detector with enhanced detection efficiency for the remaining micro-beams. Such masking layers may be placed in front of the object to be examined, between the object and the detector, or be designed as part of the detector structure itself. This present technology includes systems for x-ray microscopy using an array of micro-beams having a micro- or nano-scale beam intensity profile to provide selective illumination of micro- or nano-scale regions of an object. Each micro-beam is separated from other micro-beams by regions of lower x-ray intensity, ranging from 0.8× to 0× of the intensity of the micro-beam. An array detector is positioned such that each pixel of the detector only detects x-rays corresponding to a single micro-beam, allowing the signal arising from the x-ray detector to be identified with the specific, limited micro- or nano-scale regions illuminated. In some instances, the object being imaged and the detector are positioned within the same Talbot diffraction order. In the present system, the spatial resolution is decoupled from the source size and the detector pixel size. Imaging using Talbot fringes typically involves a grating (often a phase-shifting grating) to produce the Talbot interference pattern, and then analysis of the resulting pattern with a second grating and/or an array x-ray detector. FIG. 1A illustrates a Talbot interference fringe generated by an absorption grating G having a 50/50 duty cycle with a pitch p when illuminated by a plane wave. Interference fringes are generated behind the grating, reconstructing the pitch p with a 50/50 duty cycle at the Talbot distance DT, given by D T = 2 ⁢ p 1 2 λ [ Eqn . ⁢ 1 ] where p1 is the period of the beam splitting grating and λ is the x-ray wavelength. As an x-ray illuminator, the Talbot interference pattern can, with suitable selection of a beam-splitting grating, produce bright anti-nodes with corresponding micron-scale dimensions. For x-rays with an energy of 24.8 keV and an absorption grating with a 50/50 duty cycle and a 1 micron pitch, the Talbot distance is DT.=4 cm. The scales for the x- and y-directions of the fringes in the illustration of FIG. 1 are quite different, and although the fringes may laterally (i.e. perpendicular to the direction of propagation) have a micron scale and pitch, they can have depth-of-focus on the scale of hundreds of microns to even centimeters. Fringe patterns at various fractional Talbot distances may actually be smaller than the size of the original grating features. These anti-nodes may therefore serve as the multiple micro-beams used for illuminating an object to achieve higher resolution. The range (depth-of-focus) over which the anti-node maintains its finest dimension is related to the pitch p of the Talbot fringes by: DOF ∝ p 2 2 ⁢ λ [ Eqn . ⁢ 2 ] The waist, or “depth-of-focus” equivalent for the anti-node for x-rays of, for example, 20 keV and a grating period of 1 micron is on the order of centimeters. FIG. 1B illustrates an enlarged portion of an anti-node of FIG. 1A, with a portion that may be considered a DOFs of one of the anti-nodes noted. In some instances, an anti-node is a portion of the beam that differs from a node by greater than 20%; for example, the contrast ratio between an “anti-node” and “node” can be 1.2:1. Exact definitions of a beam “waist”, defined by the range over which an anti-node varies by less than a predetermined amount (e.g. a length range over which the anti-node full-width at half-maximum variation is within 5%) may be defined for various Talbot patterns. Note that a given interference pattern may have many fine “waists” that can be used for illumination, and, depending on the grating used, some may be of even finer dimensions than the grating half-pitch. These “waists” may also occur at any number of distances from the grating and need not be at the previously defined fractional Talbot distances. The pattern of Talbot fringes therefore resembles an array of “micro-beams” propagating in space. The fringes may be parallel micro-beams, as was illustrated in FIG. 1, or may be obtained using converging or diverging x-ray beams. Additional examples of Talbot interference patterns are shown in FIGS. 2A-2C. FIG. 2A illustrates the intensity pattern produced by a grating 210-1-90 (shown in cross section) introducing a π/2 radian phase shift from a 1:1 grating-to-space width ratio. FIG. 2B illustrates the intensity pattern produced by a grating 210-1-180 introducing a n radian phase shift in a 1:1 grating-to-space width ratio. FIG. 2C illustrates the intensity pattern produced by a grating 210-3-180 introducing a n radian phase shift in a 1:3 grating-to-space width ratio. Simulations of FIGS. 2A-2C assume gratings with a Ronchi (e.g. line/space square wave) profile and a point radiation source with sufficient spatial coherence. FIG. 2D illustrates two-dimensional phase gratings and self images for phase grating periods of π and π/2. As shown in FIG. 2D, the n period grating is in the form of a checkerboard and produces a “mesh” self image. A grating with π/2 period also has checkerboard form but produces a checkerboard self-image with inverted contrast. The x-ray microscope of the present technology can utilize a grating with a period of π, π/2, or other period to produce microbeams. In many embodiments, this beam splitting diffraction grating is that of a phase grating of low absorption but producing considerable x-ray phase shift of either π/2 or π radians, or some other specified or predetermined value such as a fraction of or multiple of π or π/2. These gratings may be one-dimensional or two-dimensional. In some embodiments, the object being examined is placed downstream of the diffractive grating at a fractional Talbot distance DN represented by the equation D N = N a ⁢ p 1 2 8 ⁢ λ = N a 16 ⁢ D T [ Eqn . ⁢ 3 ] where p1 is the period of the beam splitting grating, DN is the fractional Talbot distance for a plane wave illumination, λ is the mean x-ray wavelength, and Na is the Talbot fractional order (N=1, 2, 3, . . . ) at which the object is placed. In some instances, the object is placed downstream of the diffractive grating at a distance that is not a fractional Talbot distance, but instead located at a distance wherein the wavefront is comprised of regions of anti-nodes and nodes that correspond to the periodic regions of interest for analysis. Depending on the grating parameters (e.g. a π phase shifting grating versus a π/2 phase shifting grating), optimal Talbot distances (Na) may be chosen for interference patterns of interest or best suited for the application. 1. Talbot Fringes as an Array of Micro-Beams. The microscope system and method of using it disclosed herein may be formed using any number of techniques that create an array of micro- or nano-scale x-ray beams used for illuminating an object. As an example, using an optical system to image either multiple arrayed x-ray sources or alternatively, an x-ray source having a transmission target with an array of microstructures, may provide “micro-beams” that correspond to the images of the source points within the depth of focus of the x-ray optical system. Talbot fringes, especially those formed by a phase grating, are a highly efficient method of directing x-rays into a effective array of micro-beams. The effective lateral dimension of the Talbot anti-nodes (the beam diameter if the beams are constructed to be circular) can, using the appropriate beam-splitting grating to establish the fringes, be made to be very small (e.g. submicron, such as 20 nm or 300 nm). The Talbot interference pattern, when used to illuminate an object under investigation in transmission, provides an array of discrete micro- or nano-probes that can be detected and analyzed using an array detector. In this way, the x-ray microscope system can achieve submicron (e.g. 0.3 um) spatial resolution at high throughput. When the detector is selected to have a pixel size that corresponds to the pitch of the Talbot fringes, and both the object and the detector are placed within the effective “depth-of-focus” of the Talbot fringes, each pixel is detecting transmitted x-rays from a single one of the “micro-beams.” The contrast between the intensity of the plurality micro-beams and the regions between the micro-beams may be further improved by placing an absorbing grating of the same pitch as the micro-beams such that the x-rays between the micro-beams are attenuated. As in the previously mentioned co-pending US Patent Applications and US Provisional Patent Applications, scanning the object in x- and y-dimensions allows the micro- or nano-scale probe to be moved over the object, and if the range of motion is as large as, or larger than, the Talbot fringe pitch, a high resolution “map” of the transmission of the object may be obtained with a relatively lower resolution x-ray pixel array detector. The “resolution” of the system is dictated solely by the size of the micro-beam, and is independent of the detector pixel size. A schematic for such a system is illustrated in FIG. 3A, and in more detail in FIGS. 4A and 4B. Source 011 provides electrons 111 to target 100 to generate an x-ray beam 888 which creates an array of microbeams after passing through a grating G1. The source of X-rays satisfies known constraints to realize the arrays of beamlets, preferably down to sub-micron size. The source of X-rays can be a single point or line source, or a periodic structured source such as a conventional source paired with an absorption (one- or two-dimensional) grating. Alternatively, a key development that yields increased throughput is decoupling the source size from the spatial resolution, which allows a large and consequently high power source to be used. One innovation of the present technology that enables greater x-ray power employs an x-ray source patterned according to a periodic pattern A0. Such a system is illustrated in FIG. 3A. In this configuration as illustrated, an x-ray source 11 has a target 100 having a substrate 1000 and a region 1001 containing discrete microstructures 700 of element size a arranged in a periodic 2-D pattern with period p0. When bombarded with electrons 111, these produce x-rays 888 in a periodic pattern with period p0. In some instances, The target 100, which may include x-ray generating microstructures, x-ray blocking masks, and/or other elements described herein, can implement an x-ray generator. Each microstructure 700 in target 100 of the structured source acts as an independent and mutually incoherent sub-source (or source points) of x-rays. The interference of these source points create a set of fringes in the sample plane that are laterally displaced with respect to the other source points. The pitch of the structured source and source to G1 distance can be selected to ensure that the fringes overlap in the sample plane. The increase in focused flux is proportional to the number of source points used. In some instances, the source is sufficiently far from the G1 grating 210-2D to have a coherence length larger than the G1 grating period. If an individual sub-source apparent width is S, the distance between the source and G1 is Z and the radiation wavelength is L, then it holds that L*Z/S>p1 where p1 is the G1 period. When an array of x-ray illumination beams (micro beams) 888-M is formed, the object 240 to be examined is illuminated at an array of discrete interaction locations 282. In many embodiments, the sample 248 is placed at a Talbot distance downstream of the beam-splitting grating. The positions can be scanned in x- and y-dimensions perpendicular to the direction of propagation of the micro-beams using a position controller 245, and the x-ray illumination beams 889-T resulting from the interaction of the micro-beams and the object can be detected by an array detector 290. The array detector 290 will be aligned such that each pixel of the detector will be positioned to collect only x-rays corresponding to a single micro-beam. This is typically within the “depth of focus” of the anti-node. By pairing the use of multiple micro-beams paired with a detector having a pixel pitch matched to the pitch of the micro-beams, and aligned so that each pixel detects x-rays from only the interaction of a single micro-beam at a given position on the object, the equivalent of 102 to 104 parallel micro-beam detection systems can be created. The object can then be scanned in x- and y-coordinates. This produces “maps” in parallel of the properties of the object, but the range of motion can be reduced to only correspond to the pitch of the micro-probes (although some overlap between scanned areas may be appropriate to provide a relative calibration between data collected for neighboring “maps”). The data in each point in the map is limited in resolution only by the lateral dimensions of the Talbot fringe, so a less expensive and/or more efficient detector with larger pixels can be used to collect high resolution images. The “maps” generated by each pixel may then be stitched together digitally to produce a large-scale “macro-map” of the object properties, while reducing the corresponding data collection time by a factor related to the number of micro-beams (e.g. up to a factor of 104). To achieve some degree of tomographic analysis, limited angle adjustment of the object may also be added to the motion protocol, as long as the interaction of x-rays with the object as well as the corresponding detector pixel both remain within the depth-of-focus for all of the multiple micro-beams. 1.1 Alternative X-Ray Sources In some instances, the x-ray source target may comprise a microstructured mask. FIG. 3B illustrates a substrate 1000 with an embedded microstructure mask. The substrate 1000 of FIG. 3B includes a thin film 1002, a first substrate portion 1004, and a second substrate portion 1006. The substrate portions 1004 and 1005 may be formed of low atomic element materials such as diamond, Be, sapphire, etc. An electron beam bombarding the thin film 1002 generates x-rays within the thin film. The generated x-rays are blocked by microstructures 700 to create an effective array of x-ray sub-sources. Microstructures 700 may be placed onto substrate portion 1004 and covered or encapsulated by substrate portion 1006. Alternatively, they may be formed by embedding the microstructures within a single substrate portion, as shown in target 1000 of FIG. 3C. Though only one pattern of a microstructure element in target 1000 is illustrated in FIG. 3A-3C, other implementations are possible and considered within the scope of this disclosure. For example, target 1000 can include multiple target patterns formed by any combination of microstructures and masks, wherein one or more of the multiple target patterns can have multiple depths within a substrate. In some instances, the electron beam may be incident onto the target at an oblique angle. FIG. 3D illustrates a system having one or more electron beams 11 bombarding a target 1000 at an oblique angle, such as between 20 degrees and 80 degrees. In some instances, the incidence angle of the electron beams on the target may be about 60 degrees. Providing the incident electron beam at an oblique angle allows for a higher energy x-ray beam from the target and reduces scattering in substrates such as diamond. FIG. 3E illustrates a target having a substrate 1004 (typically a low atomic material such as diamond) and microstructures 700. In some instances of the present technology, the thickness t of the targets can be optimized for the particular material to improve contrast between x-rays that are emitted the microstructures 700 and x-rays generated in the substrate. In some cases, the thicknesses are on the order of 2-10 um. In some instances, the depth of the target microstructure material within a substrate may be optimized to achieve a particular acceleration voltage. FIG. 3F illustrates a plot of optimal thickness vs. acceleration voltage for molybdenum (Mo) microstructures. As shown, the relationship between the optimal depth in micrometers to the acceleration voltage in kilovolts is approximately linear. For example, for an energy of 60 kV, the optimal depth would be about 10 microns. Though only data for molybdenum is displayed, the optimal depth of a target microstructure for other materials may also be optimized for a particular acceleration energy. Some microstructured targets may furthermore comprise electrically conductive layers, layers to improve thermal conductivity between the microstructure and the substrate, and/or diffusion barriers. 1.3 X-Ray Source Filtering In embodiments in which the micro-beams are generated by the Talbot effect, the bandwidth of the x-ray beams at the object to be examined must be within +/−15% of a predetermined x-ray energy of interest. This is typically achieved through the use of filters, such as thin metal foils. 2. Geometric Conditions Returning to FIG. 3A, the x-rays 888 that emerge from the arrayed source as an array of individually spatially coherent but mutually incoherent sub-sources of illumination for the beam splitting grating G1 210-2D placed at a distance L from the arrayed x-ray source A0. The position of the object 240-W to be illuminated by the array of micro-beams is placed at a further distance D from the beam-splitting grating G1 210-2D. To ensure that each x-ray sub-source in A0 contributes constructively to the image-formation process, the geometry of the arrangement should satisfy the condition: p 0 = p 2 ⁢ L D [ Eqn . ⁢ 4 ] When the condition is met, the x-rays from the many sub-sources of A0 produce the same (overlapping) Talbot interference pattern, and because the various mutually incoherent sources do not interfere with each other, these Talbot patterns will add as intensities. The effect at the object 240-W is therefore to simply increasing the intensity of the micro-beams (along with it the signal-to-noise ratio) above what a single coherent source can provide. This configuration is called the Talbot-Lau interferometer. It should be noted that the arrayed x-ray source may also be provided in some embodiments using a uniform x-ray material and a masked grating that allows x-rays to emerge only from specific points arranged in an array of dimension a and period p0. An arrayed x-ray source may also be provided by selective bombardment of an x-ray generating material using a patterned electron beam. The beam-splitting grating may be an amplitude grating with a 50/50 duty cycle, as illustrated in FIG. 3A, or may be an amplitude grating with some other duty cycle. A phase-shifting beam-splitting grating may comprise a 1-D or 2-D periodic pattern of n or π/2 phase-shifts. To ensure that the object 240-W to be examined is illuminated by a periodic pattern of x-ray micro-beams, the distance D between the grating and the object should correspond to one of the fractional Talbot distances, i.e. D = n ⁢ 2 ⁢ p 1 2 16 ⁢ λ [ Eqn . ⁢ 5 ] where n is a non-zero integer. The suitable value of n may be different if the grating is a transmission grating, a n phase-shifting grating, or a π/2 phase-shifting grating. Another equation often used in Talbot-Lau systems relates the pitch p1 of the Talbot grating G1 to the size a of the x-ray generating elements in the arrayed source: p 1 ≥ L ⁢ λ a [ Eqn . ⁢ 6 ] Most embodiments of the invention employ a interferometric system in which the conditions presented in Eqns. 4-6 are met. In some embodiments, the object 240-W to be examined may be mounted on a position controller 245 that may be controlled to translate the object 240-W in x- and y-dimensions. For some embodiments, additional rotation of the object for generating tomographic imaging data may also be controlled by the mounting system. In some embodiments, a 5-axis mount, or a goniometer, may be used. It should be noted that these embodiments as illustrated are not to scale. 3. Detector Considerations As disclosed here, the detector pitch will be matched to the pitch of the multiple microbeams so that each pixel is positioned to only detect x-rays emerging from the interaction of the object with a single micro-beam, and the cross-talk between pixels due to neighboring micro-beams is minimized. Then, the data collection and final reconstruction of the “map” of the properties of the object may proceed, knowing that the distinct signals from each pixel need not be further deconvolved. If there is cross-talk between micro-beams and pixels (e.g. due to scattering or fluorescence), additional image analysis may be able to remove some of the cross-talk if it can be properly calibrated. Energy resolving array detectors may also be used to separate signals from transmitted x-rays, scattered x-rays, and fluorescence x-rays. This matching is most straightforwardly achieved if the detector pitch is a 1:1 match to the pitch of the micro-beams, i.e. each beam has a corresponding single pixel in the detector, and the detector is placed in close proximity to the object and the micro-beams. 3.1 Finer Detector Pitch In some embodiments, detector pitches that are integer fractions of the pitch of the micro-beams (e.g. a 2× reduction in pitch, which would indicate, for example, in a 2-D array, that 4 pixels are positioned to collect the x-rays corresponding to a single micro-beam, or a 3× reduction in pitch, which would indicate 9 pixels are present to detect the x-rays corresponding to each micro-beam) may also be used. This may offer some advantages if the x-rays being detected have some spatial structure, for example if the desired x-ray signal is related to small-angle scattering from the object. Then, certain pixels of the detector can be aligned to detect only the scattered x-rays, while the non-scattered beam may be collected by a different pixel, or simply blocked. 3.2. Larger Detector Pitch. In other embodiments, a larger detector pixel may be used. In this case, a pixel size that is larger than the pitch of the Talbot fringe may be used, as long as the active area of each pixel of the detector (the portion converting x-rays into an electronic signal) is on the order of the same size as the corresponding x-ray micro-beam. The detector may therefore be less expensive, and yet still produce a “high resolution” signal (since the spatial resolution is determined by the interaction volume of the Talbot fringe and the object, not the detector pixel size). One disadvantage of this technique is that only 1 out of 4 Talbot fringes is used for detection, and the other fringes are wasted. Although certain Talbot fringes will end up not being used, the missing information may still be provided by scanning over the distance between detector pixel centers. And furthermore, with a larger pixel, greater detection efficiency may be achieved for the micro-beams that are detected. FIGS. 5-12 illustrate the use of larger pixels in some embodiments of the invention. FIG. 5 illustrates a schematic of an embodiment of a system similar to that of FIG. 3A, but in which a mask has been placed in front of the object 240-W to block a certain number of micro-beams. As illustrated, 3 out of every 4 micro-beams are blocked, with only 1 beam out of 4 proceeding to illuminate the object and then be detected by the detector, but any number of beams may be blocked according to predetermined patterns for various applications. FIGS. 6A and 6B illustrate such an embodiment in more detail, presenting illustrations similar to those of FIGS. 4A and 4B. As can be seen by the comparison with FIGS. 4A and 4B, because only a certain number of micro-beams are used, the pitch of beams at the detector is substantially larger, and a less expensive detector with a larger pixel size may be used. As illustrated in FIGS. 3 through 6B, the x-ray detector is presented as a direct array detector, generating an electrical signal in response to the absorption of x-rays. Such an electronic sensor may directly create an electrical signal in response to the absorption of x-rays, by, for example, the creation of direct electron-hole pairs in amorphous selenium (a-Se). These are then converted into electronic signals using an array of thin-film transistors (TFTs). Such direct flat panel detectors (FPDs) such as the Safire FPD of Shimadzu Corp. of Kyoto, Japan, are commercially available. In other embodiments, the detector may use scintillators that emit visible or ultraviolet light when exposed to x-rays. The active x-ray detection region may be defined, for example, by providing a scintillator such as cesium iodide doped with thallium (CsI(Tl)) or by providing a detector with a uniform coating of scintillator with a masking layer of high Z material, for example, gold (Au), on top. FIG. 7 illustrates a variation of the embodiment of FIG. 6B, but using a detector 290-S in combination with a fluorescent screen or scintillator 280. The scintillator 280 comprises a material that emits visible and/or UV photons when x-rays are absorbed, and the detector 290-S detects those visible and/or UV photons. Typical scintillator materials comprise a layer of cesium iodide (CsI), thallium doped CsI, yttrium aluminium garnet (YAG) or gadolinium sulfoxylate (GOS). In conventional imaging systems, high resolution images with a scintillator-type detector in close proximity to the object can be obtained, but the overall thickness of the scintillator and electronic elements must be thin enough so that each detector pixel is collecting only x-rays corresponding to that pixel. However, in the system disclosed herein, the spatial resolution is defined by the dimensions of the micro-beams 888-M instead of the detector pixel size. This allows a larger pixel and therenby a thicker scintillator material to be used, since every photon generated from the larger pixel will be known to have originated from a predetermined micro-beam. The thicker scintillator increases the probability that a given x-ray photon will be absorbed and converted to visible light, increasing the potential signal. Some additional number of x-ray photons will generate secondary electrons in the scintillator material, which may in turn excite additional visible/UV emission from the scintillator material. However, as all x-ray photons within the pixel are known to have originated from a single micro-beam, the additional photons emerging from this excitation are also known to have their origin with these spatially defined x-rays, and simply increase the overall signal that may be detected. FIG. 8 illustrates an additional variation on a system using a scintillator, in which the visible/UV light 890 from the scintillator 280 is collected by a visible/UV optical system 320 and imaged onto a detector 290-SI. The visible/UV optical system may comprise optics with additionally magnify the image of the scintillator. When using relay optics and a magnified image, the electronic detector need not comprise a high resolution sensor itself, and less expensive commercial CCD detectors or complementary metal-oxide-semiconductor (CMOS) sensor arrays with, for example, 1024×1024 pixels, each 24 μm×24 μm square, may be used. Thicker scintillators may also be used in some embodiments having relay optics, increasing sensitivity. However, when relay optics are used, detection is limited to the field of view collected by the x-ray optics, which may in some cases be only on the order of hundreds of microns. Collecting data on larger areas will then need to be “stitched” together from several exposures. FIGS. 9, 10A and 10B represent an additional embodiment in which a masking structure 297 is placed between the object 240 and the detector. For this embodiment, all available micro-beams 888-M illuminate the object 240, but a masking layer 297 made of, for example, gold (Au), prevents 3 out of every 4 beams from entering the detector 290. This also allows detector 290 to have a larger pixel, again reducing cost for direct detectors and, for embodiments using scintillators, increasing potential detector efficiency. FIG. 11 illustrates an additional variation of the embodiment of FIGS. 9, 10A and 10B, but with the detection of x-rays achieved using a scintillator 280 and a visible/UV detector 290-S. FIG. 12 illustrates an additional variation on a system using a scintillator, in which the visible/UV light 890 from the scintillator 280 is collected by a visible/UV optical system 320 and imaged onto a detector 290-SI. Although the scintillators as illustrated in FIGS. 7, 8, 11, and 12 are shown as comprising uniform layers of scintillator, embodiments using patterned scintillator material, in which scintillator material is placed only over a portion of the pixel, may also be used. The selective placement of scintillator material over portions of the detector may be used as an alternative to the use of a masking layer to select certain micro-beams for detection. Detectors with additional structure within each pixel may also be employed as well. For example, if the typical detector pixel is 2.5 microns by 2.5 microns (an area of 6.25 micron2), but the micro-beam diameter is only 1 micron, a detector pixel with a central “spot” of scintillator material slightly larger than 1 micron, surrounded by “dead” zones, and positioned to correspond to the position of the micro-beam may be created. With this configuration, all the x-rays from the micro-beam should be detected, while reducing the detection of scattered or diffracted x-rays that would otherwise cause spurious signals if the full area of the detector pixel were to be used. Likewise, pixels in which detector structures (such as scintillator material) are only positioned on the outer portion of the pixel, for example, to only detect x-rays scattered at small angles while not detecting the directly transmitted beam, may also be used for some embodiments. Likewise, although the mask 297 in FIGS. 11 and 12 is shown as displaced from the scintillator layer, some embodiments may have the masking layer directly deposited onto the scintillator layer. Other embodiments for patterned scintillators may be known to those skilled in the art. 3.0 Methods of Microscopic Data Gathering. FIG. 13 illustrates method for collecting microscopy data. The data collection may be used to form a 2-D “map” or 3-D tomographic image. X-ray microbeams are generated in step 4210 through the use of an x-ray source and a beam-splitting grating, preferably a phase grating. In some instances, the x-ray source employs an x-ray target comprised of microstructures on or embedded within a substrate of low mass density (e.g. diamond or Be). In some instances, the x-ray source employs a target comprising a thin film coated on top of a substrate of low mass density and furthermore comprising embedded microstructures that serve as a “mask” to block a portion of the x-ray beams. In some instances, the x-ray source is an extended x-ray source and is used in combination with an absorbing grating. In some instances, the x-ray source is a microfocus x-ray source. A filtering method is placed 4220 between the x-ray source and the beam-splitting grating to limit the bandwidth of the x-rays from the x-ray source to a bandwidth. In some instances, the bandwidth of the illumination beam can be ±15%, depending on which pre-determined Talbot or fractional Talbot distance is used. An object to be examined is aligned 4230 at a Talbot distance such that the region of nodes (darkest intensity) and anti-nodes (highest intensity) of the microbeam has a pitch p in the directions orthogonal to the propagation direction (designated the “x” and “y” directions) is 20 micrometers or less. The contrast between regions of greatest intensity (generally at the center of the micro-beams) and the darkest intensity (generally the region exactly between micro-beams) is preferred to be at least 20%, although in some cases, an intensity ratio of 1.2:1 or 2:1 between the anti-nodes and nodes may provide enough contrast. In some instances, the bandwidth of the illumination beam satisfies the following equation: Δλ = λ 0 2 ⁢ m - 1 . A detector is aligned 4240 within the “waist” of the microbeams so that each detector pixel generates signals corresponding to a single microbeam. For the micro-beams formed by an imaging system, this position may correspond to the depth-of-focus of the imaging system. In most instances, the detector pixel pitch and microbeam are the same or approximate with some scaling, such that the center of each microbeam is coincident upon the center of the detector pixel. For micro-beams formed by a Talbot system, this may correspond to the position of the interference pattern at a fractional or integer multiple of the Talbot Distance, where self-replicating images are formed. There is some flexibility in the exact positioning of the detector, as long as each pixel of the detector generates a signal corresponding only to a single micro-beam (without cross-talk between the micro-beams or detector pixels). Generally, a detector will be chosen where every micro-beam has a corresponding pixel or set of pixels; however, in some embodiments, the detector may only detect a subset of the micro-beams. In some instances, a detector can be chosen to having a pixel pitch pd equal to a non-zero integer multiple of the micro-beam pitch p. X-rays transmitted by each microbeam are recorded 4250 by the detector, and the corresponding electronic signals representing x-ray intensity and energy are recorded. If only a single set of datapoints are desired, no more data need be collected. In most embodiments, however, the object to be examined is moved 4260 using a position controller to build up a 1-D or 2-D “map” of the properties of the object. This is typically performed so that the object is moved several times corresponding to to the FWHM of each microbeam region of highest intensity and moved in both x and y dimensions. If no information beyond a 2-D scan in x- and/or y-dimensions is needed, the present system can take the accumulated data and, in this case, use various image “stitching” techniques that are generally well known in the art, synthesize a 2-D intensity “map” representing the large-area x-ray transmission/absorption of the object. If, on the other hand, 3-D information is desired, the object is rotated through an angle relative to the z-axis (this rotation may be a rotation around either the x- or y-dimensions) to collect a set of data from the x-ray detector at this alternative rotation position. The system will loop through these steps to collect x-ray information at a preprogrammed sequence of positions and rotations until a complete set of data is collected. At this point, the system will then proceed to take the accumulated data and, in this case, use various image 3-D analysis techniques that are generally well known in the art, to synthesize a 3-D representation of the large-area x-ray transmission/absorption of the object. Variations on the method described above may also be put into practice. For example, instead of first executing a loop of data collection in x- and y-dimensions at a fixed rotation position, and then changing the rotation setting to collect additional data, embodiments in which the object is rotated by a mechanical mechanism while the x- and y-position settings remain fixed may also be executed. Rotation of the object around the z-axis may also provide additional information that can be used in image tomosynthesis. 4. Limitations and Extensions. With this Application, several embodiments of the invention, including the best mode contemplated by the inventors, have been disclosed. It will be recognized that, while specific embodiments may be presented, elements discussed in detail only for some embodiments may also be applied to others. Also, details and various elements described as being in the prior art may also be applied to various embodiments of the invention. While specific materials, designs, configurations and fabrication steps have been set forth to describe this invention and the preferred embodiments, such descriptions are not intended to be limiting. Modifications and changes may be apparent to those skilled in the art, and it is intended that this invention be limited only by the scope of the appended claims.
	abstract	The invention relates to a device limiting the appearance of decoding artefacts for a gamma camera with a coded mask comprising a gamma radiation detector (3) opposite the coded mask (1) and having a field of view with an area partially coded (20) by the coded mask (1). It comprises a recessed part (30), which is opaque to the gamma radiation, to be arranged opposite the detector (3) with respect to the coded mask (1), with said recessed part (30) obscuring the partially coded area (20) of the field of view for the detector (3).
050193279	abstract	A fuel assembly transfer basket for a pool type, liquid metal cooled nuclear reactor having a side access loading and unloading port for receiving and relinquishing fuel assemblies during transfer.
051241137	abstract	The invention relates to a nuclear reactor with improved efficiency.. The nuclear reactor comprises a massive partition (18) of material reflecting high energy neutrons at the circumference of the core (7) of the reactor, two layers (19 and 20) of material absorbing low energy neutrons and containing fertile material arranged one at the lower part and one at the upper part of the core (7) and an assembly of neutron energy spectrum variation rods (27). The rods (27) are associated with mechanisms permitting them to be either fully inserted into the core (7), or fully extracted. These rods (27) consist of a material absorbing low energy neutrons and permit a shift of the neutron spectrum towards the high energies.. The invention is particularly applicable to pressurized water nuclear reactors.
048719137	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to a signal processing method in autoradiography for determining base sequence of DNA or DNA fragment. 2. Description of the Prior Arts Autoradiography has been known as a method for obtaining locational information on radioactively labeled substances distributed in at least one-dimensional direction to form distributed rows on a support medium. For instance, the autoradiography comprises steps of: labeling organism-originating biopolymers such as proteins or nucleic acids with a radioactive element; resolving the radioactively labeled biopolymers, derivatives thereof, cleavage products thereof or synthetic products thereof on a gel support (support medium) through a resolving process such as gel electrophoresis; placing the gel support and a high-speed X-ray film together in layers for a certain period of time to expose said film, developing said film; obtaining the locational information on the radioactively labeled substances from the developed film; and then performing the identification of the polymeric substances, determination of molecular weight thereof and isolation thereof based on the obtained locational information. Recently, the autoradiography has been effectively used especially for determining the base sequence of a nucleic acid such as DNA (or DNA fragment, hereinafter "DNA" may be used to include both DNA and DNA fragment). The autoradiograhy is considered to be of a value for the structural determination of polymeric substances originating from organisms. Sanger-Coulson method is known as a representative method for determining the base sequence of DNA. In this method, the base sequence of DNA is determined by ingeniously utilizing synthesis of DNA fragments with DNA synthesis enzyme, gel electrophoresis and autoradiography on the basis of such characteristic structure that DNA has a double helix structure consisting of two chain molecules, whose constitutional unit contains any one of four bases: adenine (A), guanine (G), cytosine (C) and thymine (T), the two chain molecules are stabilized through hydrogen bonding between these bases, and that the hydrogen bonding between each two constitutional base units comprises only two combinations, namely, G-C and A-T. In the Sanger-Coulson method, there are a number of known procedures for synthesizing and sequencing DNA fragments complementary to DNA (hereinafter the sequenced fragments being referred to as DNA specimen). Basically, single-stranded DNA is used as template, which is incubated with a DNA synthesis enzyme (DNA polymerase) in the presence of mononucleoside triphosphates containing the above four kinds of bases, thus synthesizing (transcribing) DNA fragments of different but discrete lengths which are complementary to the DNA specimen. When part of the mononucleoside triphosphates are radioactively labeled, there can be obtained base-specifically synthesized DNA fragments (synthetic DNAs) labeled with a radioactive element. A mixture of the synthetic DNAs is then resolved (developed) on a support medium by electrophoresis, and the support medium is autoradiographed to obtain an autoradiograph of a resolved pattern of the synthetic DNAs. The base sequence can be determined in order, starting from the end of the chain molecule according to the visualized autoradiograph. In this way, the sequence for all bases of the DNA specimen can be determined. Sanger-Coulson method summarized above is described briefly in the following literature: "Reading the genetic information in the original language. A surprising method for sequencing the bases of DNA" written in Japanese by Kin-ichiro Miura, Modern Chemistry, September 1977, pp. 46-54 (Tokyo Kagaku Dozin Ltd., Japan). In the autoradiography utilizing the conventional radiography, as described above, the visualization of an autoradiograph on a radiographic film is needed to obtain locational information on radioactively labeled substances. In fact, the investigators have determined the distribution of radioactively labeled substances in a sample by visually observing a visualized autoradiograph to obtain the locational information on the specific substances labeled with a radioactive element. For instance, the DNA sequencing is visually determined by measuring the resolved positions of base-specific synthetic products labeled with a radioactive element and comparing them with each other. Accordingly, the autoradiography has unfavorable feature that it requires visual analysis, thereby needing a great amount of time and labor. Further, the locational information obtained by the analysis of the autoradiograph varies depending on the investigators in charge because of inherent unreliability of visual observation, and the accuracy of the information is limited to a certain extent. Particularly, in such a case that only a small quantity of a sample has been employed, that the intensity of radiation energy emitted by the radioactively labeled substances has been low, or that the exposure has been done under unfavorable conditions, an autoradiographic image visualized on a radiographic film has reduced quality (in regard of sharpness, contrast, etc.) so that satisfactory information can not be obtained and the accuracy is low. These are drawbacks of the conventional autoradiography. In order to improve the accuracy of the locational information, for instance, a visualized autoradiograph can been scanned by means of a device such as a scanning densitometer. However, such scanning process requires increased time and complicated procedures. In addition to the above-described disadvantages, a sample and radiographic film is required to be placed together in layers for a long period of time to expose the film to radioactive substances contained in the sample so as to visualize the autoradiograph having the locational information on the radiographic film (for instance several days are required). Moreover, the exposure ought to be conducted at a low temperature (0.degree. C. to -90.degree. C.) to avoid appearance of the chemical fog of the photosensitive silver salt in the radiographic film which is caused by various substances contained in the sample. Thus, the exposure ought to be done under specific conditions. Moreover, since the photosensitive silver salt in a radiographic film has a drawback that it is also sensitive to physical impetus, or is apt to be physically fogged, careful and skillful handling is required. This feature further brings about another complexity in the autoradiographic procedure. The radiographic film is also exposed to certain natural radioactive substances contained in the sample other than the radioactively labeled substances under analysis during the exposure for a long period of time. Accordingly, this exposure also lowers the quality of the obtained image, in addition to the chemical and physical fogs do. SUMMARY OF THE INVENTION The present inventors have found that the base sequence of DNA or DNA fragment is easily determined by subjecting digital signals which corredspond to an autoradiograph having locational information on the radioactively labeled substances to an appropriate signal processing. Said digital signals can be obtained without necessity of visualization thereof by utilizing a radiation image recording and reproducing method using a stimulable phosphor sheet. Further, the visualization of the autoradiograph based on the obtained electric signals or digital signals can be attached to the signal processing. The present invention provides a signal processing method in autoradiography for determining base sequence of DNA or DNA fragment, employing at least four groups of base-specific synthetic products which are complementary to the DNA or DNA fragment and labeled with a radioactive element, said at least four groups comprising: (1) base-specific synthetic products including at least guanine-specific synthetic products; (2) base-specific synthetic products including at least adenine-specific synthetic products; (3) base-specific synthetic products including at least thymine-specific synthetic products; and (4) base-specific synthetic products including at least cytosine-specific synthetic products; and being resolved one-dimensionally in parallel relation to each other to form resolved rows on a support medium, which comprises a process including: (1) determining on each of the resolved rows a scanning line for signal processing; (2) detecting on each of the resolved rows sampling points on said scanning line; and (3) comparing and identifying the positions of said sampling points on the scanning lines among the resolved rows to obtain locational information on guanine, adenine, thymine and cytosine; said process being applied to digital signals corresponding to an autoradiograph having the locational information on the radioactively labeled synthetic products, said digital signals being obtained by causing a stimulable phosphor sheet to absorb radiation energy emitted by the radioactively labeled synthetic products in the support medium to record the autoradiograph of the radioactively labeled synthetic products on the phosphor sheet, irradiating said phosphor sheet with an electromagnetic wave to release at least a portion of the radiation energy stored in the phosphor sheet as stimulated emission, and detecting the stimulated emission photoelectrically. A visible image can be further obtained by attaching to the signal processing method an additional process for reproducing the visible image from electric signals or digital singals corresponding to said autoradiograph obtained by detecting the stimulated emission photoelectrically. The digital signals corresponding to the autoradiograph can be also obtained by reading out an autoradiograph visualized on a radiosensitive material according to the conventional radiography. Accordingly, the present invention also provides a signal processing method in autoradiography for determining base sequence of DNA or DNA fragment, employing the above-mentioned radioactively labeled synthetic products which are complementary to the DNA or DNA fragment and resolved one-dimensionally in parallel relation to each other to form resolved rows on a support medium, which comprises a process including: (1) determining on each of the resolved rows a scanning line for signal processing; (2) detecting on each of the resolved rows sampling points on said scanning line; and (3) comparing and identifying the positions of said sampling points on the scanning lines among the resolved rows to obtain locational information on guanine, adenine, thymine and cytosine; said process being applied to digital signals corresponding to an autoradiograph having the locational information on the radioactively labeled synthetic products, said digital signals being obtained by exposing a radiosensitive material to radiation emitted by the radioactively labeled synthetic products in the support medium to record the autoradiograph of the radioactively labeled synthetic products on the radiosensitive material, and reading out the autoradiograph photoelectrically. In the invention, the term "locational information" on the radioactively labeled substances resolved on the support medium means to include a variety of information relating to the location of the radioactively labeled substances or the aggregation thereof, being present in the support medium, such as the location, the shape, the concentration, the distribution and combination thereof.
	description	This application claims the benefit of Chinese Patent Application No. 200910118737.2 filed Jan. 24, 2009, which is hereby incorporated by reference in its entirety. The embodiments described herein relate to a filter and an X-ray imaging system and, more particularly, to a filter for adjusting the spectrum of X-ray and an X-ray imaging system provided with such a filter. In an X-ray imaging system, the spectrum of X-ray is adjusted by a filter and thereafter the X-ray is radiated to a subject. The filter is installed within a collimator box attached to an X-ray tube. In order to obtain a desired spectrum, plural filter plates of the filter, which are attached to a rotary disc, are used selectively (see, for example, Japanese Unexamined Patent Publication No. Hei 11 (1999)-76219 (Paragraph No. 0009, FIGS. 1 and 2)). It is desirable that the spectrum be adjustable in multiple stages. However, in the configuration wherein filter plates are switched over from one to another by device of a rotary disc, the adjustment is up to four stages at most. If an attempt is made to make adjustment possible in seven or more multiple stages in the rotary disc method, the rotary disc for the mounting of many filters becomes large-sized and thus such an attempt is not practical. Accordingly, embodiments of the present invention provide a filter permitting the adjustment of spectrum in multiple stages and that permitting the reduction of size, as well as an X-ray imaging system provided with such a filter. In a first aspect there is provided a filter for adjusting the spectrum of passing radiation, comprising: a support plate having an aperture for passage therethrough of radiation; a plurality of filter plates supported by the support plate and having mutually different filer characteristics; and moving device for moving the plural filter plates selectively to a position to close the aperture and a position to open the aperture. In a second aspect there is provided, in accordance with the first aspect, a filter wherein the moving device comprises: spring device for operating the filter plates in a direction away from the aperture; a plurality of ratchets plates supported by the support plate in proximity to the filter plates and each having a notched portion; a first support point for supporting the filter plates pivotably on the support plate; a second support plate for supporting the ratchet plates pivotably on the support plate; a leaf spring connected to the filter plates and the ratchet plates, causing the filter plates and the ratchet plates to pivot interlockedly with each other; and a pawl engageable with the notched portions of the ratchet plates, the pawl being moved to pivot the ratchet plates, thereby making the filter plates movable to the position to close the aperture. In a third aspect there is provided, in accordance with the first aspect, a filter wherein the support plate has a top layer and a bottom layer parallel to each other, and the filter plates are supported by the top layer and the bottom layer. In a fourth aspect there is provided, in accordance with any of the first to the third aspects, a filter wherein the support plate is in the shape of disc and is formed with the apertures at a center thereof. In a fifth aspect there is provided, in accordance with any of the first to the fourth aspects, a filter further comprising a ring capable of rotating along the circumference of the support plate, and the pawl is provided in the ring. In a sixth aspect there is provided an X-ray imaging system for radiographing a subject with X-ray passing through a filter, the filter comprising: a support plate having an aperture for passage therethrough of radiation; a plurality of filter plates supported by the support plate and having mutually different filter characteristics; and moving device for moving the plural filter plates selectively to a position to close the aperture and a position to open the aperture. In a seventh aspect there is provided, in accordance with the sixth aspect, an X-ray imaging system wherein the moving device comprises: spring device for operating the filter plates in a direction away from the aperture; a plurality of ratchet plates supported by the support plate in proximity to the filter plates and each having a notched portion; a first support point for supporting the filter plates pivotably on the support plate; a second support point for supporting the ratchet plates pivotably on the support plate; a leaf spring connected to the filter plates and the ratchet plates, causing the filter plates and the ratchet plates to pivot interlockedly with each other; and a pawl engageable with the notched portions of the ratchet plates, the pawl being moved to pivot the ratchet plates, thereby making the filter plates movable to the position to close the aperture. In an eighth aspect there is provided, in accordance with the sixth aspect, an X-ray imaging system wherein the support plate has a top layer and a bottom layer parallel to each other, and the filter plates are supported by the top layer and the bottom layer. In a ninth aspect of the present invention there is provided, in accordance with any of the sixth to eighth aspects, an X-ray imaging system wherein the support plate is in the shape of a disc and is formed with the aperture at a center thereof. In a tenth aspect there is provided, in accordance with any of the sixth to ninth aspects, an X-ray imaging system wherein the filter further comprises a ring capable of rotating along the circumference of the support plate, and the pawl is provided in the ring. According some embodiments, since the filter for adjusting the spectrum of passing radiation comprises: a support plate having an aperture for passage therethrough of radiation; a plurality of filter plates supported by the support plate and having mutually different filter characteristics; and moving device for moving the plural filter plates selectively to a position to close the aperture and a position to open the aperture, it is possible to realize a filter capable of adjusting the spectrum in multiple stages and capable of attaining the reduction of size, as well as an X-ray imaging system provided with such a filter. Embodiments of the present invention will be described in detail hereinunder with reference to the drawings. The present invention is not limited to the embodiments described herein. FIG. 1 shows a schematic configuration of an X-ray imaging system. AS shown in the same figure, this system includes an X-ray irradiator 10, an X-ray detector 20 and an operator console 30. The X-ray irradiator 10 and the X-ray detector 20 are opposed to each other with a subject 40 therebetween. The X-ray irradiator 10 includes an X-ray tube 12 and a collimator box 14. A filter 16 and a collimator 18 are accommodated within the collimator box 14. The filter 16 is an example of the best mode for carrying out the invention. With the configuration of the filter 16 there is shown an example of the best mode for carrying out the invention with respect to the filter. X-ray emitted from the X-ray tube 12 is adjusted its spectrum by the filter 16 and is then radiated to the subject 40 through an aperture of the collimator 18. The filter 16 is a spectrum variable filter. The collimator 18 is an aperture variable collimator. The X-ray which has passed through the subject 40 is detected by the X-ray detector 20 and is inputted to the operator console 30. On the basis of the inputted signal the operator console 30 reconstructs a radioscopic image of the subject 40. The reconstructed image is displayed on a display 32 of the operator console 30. The operator console 30 controls the X-ray irradiator 10. Control of the filer 16 and that of the collimator 18 are included in the control of the X-ray irradiator 10 made by the operator console 30. Where required, the filter 16 and the collimator 18 can be adjusted manually. FIG. 2 shows the configuration of the filter 16 schematically. As shown in FIG. 2, the filter 16 is of a two-layer structure comprising a top layer 100 and a bottom layer 200. The top layer 100 has an aperture 102 and the bottom layer 200 has an aperture 202. The apertures 102 and 202 are concentric and equal in size and X-ray passes through the apertures 102 and 202. The top layer 100 and the bottom layer 200 each have on the outer periphery thereof such a ring as will be described later. The rings are integrally driven by a motor 300 via a gear 302 so as to be rotatable reversibly in clockwise and counterclockwise directions. FIG. 3 shows a ring-detached state and a ring-attached state with respect to the top layer 100 and the bottom layer 200. As shown in FIG. 3, the top layer 100 and the bottom layer 200 include disc-like top support 104 and bottom support 204 assembled together in facing relation to each other, as well as rings 106 and 206 attached respectively to their outer peripheries in a rotatable manner. The top support 104 and the bottom support 204 do not rotate and are formed with the apertures 102 and 202 respectively at their centers. The rings 106 and 206 are integrally driven by the motor 300 via the gear 302 and are adapted to rotate along the circumference of the top support 104 and that of the bottom support 204 respectively. FIG. 4 shows a disassembled state of the filter 16 into the top layer 100 and the bottom layer 200. FIG. 4 shows a state in which the top support 104 and the bottom support 204 are opened to both right and left at the confronting portion thereof. The top support 104 and the bottom support 204 are coupled together by three holes 108a, 108b, 108c formed in the top support 104 and three studs 208a, 208b, 208c formed on the bottom support 104. The rings 106 and 206 are coupled together by three holes 109a, 109b, 109c formed in the ring 106 and three studs 209a, 209b, 209c formed on the ring 206. The top support 104 has three filter plates 110, 120 and 130 on an inner surface thereof, i.e., on its surface confronting the bottom support 204. The filter plates 110, 120 and 130 comprise three steel plates or the like different in thickness and supported by a support frame. For example, the three steel plates or the like are 0.1, 0.2 and 0.3 mm, respectively, in thickness. The filter plates 110, 120 and 130 constitute a part of three parallel link mechanisms, and with operation of the parallel links, the filter plates are adapted to advance and retreat to a position to close the aperture 120 and a position to open the same aperture. The three parallel links are driven in order by a pawl 140 which rotates together with the ring 106. Here there is shown a state in which with operation of the pawl 140 the filter plate 110 has advanced to the position to close the aperture 102, while the filter plates 120 and 130 have retreated to the position to open the aperture 102. The bottom support 204 has three filter plates 210, 220 and 230 on an inner surface thereof, i.e., on its surface confronting the top support 104. The filter plates 210, 220 and 230 comprise three steel plates or the like different in thickness and supported by a support frame. For example, the three steel plates or the like are 0.6, 0.9 and 1.5 mm, respectively, in thickness. The filter plates 210, 220 and 230 constitute a part of the three parallel links, and with operation of the parallel links, the filter plates are adapted to advance and retreat to a position to close the aperture 202 and a position to open the same aperture. The three parallel links are driven in order by a pawl 240 which rotates together with the ring 206. Here there is shown a state in which with operation of the pawl 240 the filter plate 210 has advanced to the position to close the aperture 202, while the filter plates 220 and 230 have advanced to the position to open the aperture 202. FIGS. 5A and 5B show the configuration and operation of parallel links with respect to the filter plate 110. Also as to the other filter plates 120 to 230, the configuration and operation of parallel links are the same. As shown in FIGS. 5A and 5B, the parallel links are made up of links A, B, C, D and joints a, b, c, d. The link A is a fixed link, the link B is a driving link, the link C is a parallel moving link, and the link D is a driven link. The driven link D is constituted by the filter plate 110 and the driving link B is constituted by a ratchet plate 112. The filter plate 110 is rotatable about the joint a, the ratchet plate 112 is rotatable about the joint b, and both are connected together at the joints c and d through the link C. With such an operation of the parallel links, the filter plate 110 assumes either such a retreated state as FIG. 5A or such an advanced state as FIG. 5B. The filter plate 110, in its retreated state FIG. 5A, opens the aperture 102, while in its advanced state FIG. 5B it closes the aperture 102. As shown in FIGS. 6A, 6B, and 6C, a tensile force induced by a spring 114 acts on the filter plate 110 constantly. Consequently, while the parallel links are not in operation, the filter plate 110 is in abutment against a stop pin 116, as shown in FIG. 6A. The parallel links operate when the ratchet plate 112 is pushed up by the pawl 140. The pawl 140 rotates while describing a circular path together with the ring 106 and pushes up the ratchet plate 112 when it rotates in the counterclockwise direction, thereby causing rotation of the ratchet plate 112. As a result, the parallel links operate and the filter plate 110 advances to the position to close the aperture 102, as shown in FIG. 6B. The engagement between the ratchet plate 112 and the pawl 140 in this state is as shown on a larger scale in FIG. 6C. As shown in FIG. 7, the pawl 140 is attached to the ring 106 rotatably with a shaft 142. A rotating force is applied in the counterclockwise direction in the figure by means of a coil spring 144 mounted on the shaft 142 and the pawl 140 is in an abutted state against a stop pin 146. One end of the coil spring 144 is fixed to the pawl 140 and an opposite end thereof is fixed to a pin 148. When the pawl 140 stays at the position b as shown in FIGS. 6A, 6B, and 6C, the filter plate 110 stays at its advanced position. As the pawl 140 further moves in the counterclockwise direction from this state, the pawl 140 passes while pushing aside the ratchet plate 112. At this time, with the restoring force of the spring 114, the filter plate 110 retreats and the aperture 102 is opened. When the ring 106 rotates in the clockwise direction, the pawl 140 cannot push up the ratchet plate 112, so that the ratchet plate 112 is not actuated and hence the parallel links do not operate. A description will now be given about a synthetic operation of the filter 16 thus constructed. Behaviors of the filter plates with rotation of the pawls 140 and 240 are shown in order in FIGS. 8 to 14. FIGS. 8 to 14 are diagrams in which the top layer 100 and the bottom layer 200 are seen in an X-ray irradiating direction. FIG. 8 shows an initial state of both top layer 100 and bottom layer 200. In the initial state, none of the filter plates close the apertures 102 and 202. This corresponds to the state in which a filter plate having a thickness of zero has been inserted into the apertures 102 and 202. With counterclockwise rotation of the rings 106 and 206 and with drive by the pawl 140, filter plates 0.1, 0.2 and 0.3 located on the top player 100 side advance in order to the position to close the aperture 102, as shown in FIGS. 9, 10 and 11. In this way X-ray spectrum is adjusted by each of the filter plates. During this period, on the bottom layer 200 side, all the filter plates stay in their retreated positions because the pawl 240 passes while stroking all the ratchet plates. With clockwise rotation of the rings 106 and 206 and with drive by the pawl 240, filter plates 0.6, 0.9 and 1.5 located on the bottom layer 200 side advance in order to the position to close the aperture 202, as shown in FIGS. 12, 13 and 14. In this way X-ray spectrum is adjusted by each of the filter plates. During this period, on the top layer 100 side, all the filter plates stay in their retreated positions because the pawl 140 passes while stroking all the ratchet plates. In this way it is possible to obtain the filter 16 whose filter plate thickness changes in seven stages. The filter 16 has a two-layer structure of both top layer 100 and bottom layer 200, and each layer has three filter plates driven by parallel links, thus facilitating reduction of the entire size. Two pawls for actuating parallel links may be provided for each layer as in FIG. 15. With two pawls 140a and 140b and 240a and 240b, it is possible to shorten the time for access to the filter plate located at the remotest position from the current position of each pawl. When the number of switching stages of filter plates may be four, the filter 16 may be of a single-layer structure of either the top layer 100 or the bottom layer 200. Further, the number of filter plates per layer is not limited to three, but may be any other suitable number.
	claims	1. An EUV radiation source, comprising: a first electrode having a first through hole; a second electrode having a second through hole; a movable insulator, having a plurality of third through holes, provided between the first and second electrodes, wherein actuating gas is introduced in the first, second through holes and one of the plurality of third through holes, and voltage is impressed between the first and second electrodes when the first, second and plural through holes are located on a common axis. 2. The EUV radiation source according to claim 1 , wherein the insulator is disciform. claim 1 3. The EUV radiation source according to claim 2 , wherein the insulator is rotated with respect to a point. claim 2 4. The EUV radiation source according to claim 2 , wherein the plurality of third through holes are concentrically provided. claim 2 5. The EUV radiation source according to claim 1 , further including an ultraviolet laser to generate pre-plasma. claim 1 6. An EUV radiation source, comprising: a first electrode having a first through hole; a second electrode having a plurality of second through holes; a movable insulator having a plurality of third through holes, each of which is connected to one of the plurality of second through holes, wherein the second electrode and the insulator are integrally provided, and wherein actuating gas is introduced in the first hole, one of the plurality of second through holes and one of the plurality of third through holes connected to the one of the plurality of second through holes, and voltage is impressed between the first and second electrodes when the first through hole, the one of the plurality of second through holes and the one of the plurality of third through holes are located on a common axis. 7. The EUV radiation source according to claim 6 , wherein the insulator is disciform. claim 6 8. The EUV radiation source according to claim 6 , wherein the insulator is rotated with respect to a point. claim 6 9. The EUV radiation source according to claim 6 , wherein the plurality of third through holes are concentrically provided. claim 6 10. The EUV radiation source according to claim 6 , wherein the first electrode and the insulator form hollow portion therein for passing through coolant. claim 6 11. The EUV radiation source according to claim 6 , further including an ultraviolet laser to generate pre-plasma. claim 6
047131993	abstract	The invention concerns a depository for radioactive waste and spent fuel cells comprising a concrete storage block in which several vertical storage and cooling channels are provided for accommodating the radioactive waste and the fuel cells. In order to provide an improved and safer concrete depository vis-a-vis the state of the art, the storage block is composed of several concrete blocks disposed beside and above one another. In the concrete blocks there are several vertical channels. The aligned channels of several concrete blocks placed on top of one another form storage and cooling channels for accommodating tubular storage containers for the radioactive waste and radioactive spent fuel cells. An annular gap is left for cooling air between the outer wall of the storage containers and the inner wall of the storage and cooling channels.
	summary
060944749	description	IDENTIFICATION OF THE COMPONENTS The following is a list of the reference numerals, along with the names of the corresponding components, that are used in the appended figures and in the description. ______________________________________ 10 gantry mounting radiography apparatus 11 first arm 12 second arm 13 vertical arm 14 table 20 compensating device 21 filter plane 22 left lung compensating filter (for internal compensation) 24 right lung compensating filter (for internal compensation) 31 axes 32 supporting plate 34 arms (of the supporting assembly) 36 hooks 40 support 42 arms (of the support) 50 X-ray source 52 X-ray tube amplifier 60 patient 62 lungs (of the patient) 63 median plane (of the organ) 64 head (of the patient) 66 T-shaped compensating filters (for external compensation) 70 star-shaped compensating filter (for external compensation) 72 flat top portion (of a T-shaped compensating filter) 74 side openings 100 cervical peripheral filter 102 neck peripheral filter 104 head peripheral filter 106 full face mask peripheral filter 108 partial face mask peripheral filter 110 straps 112 foam pillow ______________________________________ DETAILED DESCRIPTION OF PREFERRED EMBODIMENTS Gantry mounting radiography apparatus FIG. 1 is a general representation of the rotating gantry mounting radiography apparatus (10). The apparatus (10) comprises two opposite arms (11,12). The opposite arms (11,12) are connected to each other, forming a fork and are attached to a vertical arm (13) of the apparatus (10) at substantially the center of the fork. The connection point of the two opposite arms (11,12) to the apparatus (10) allows rotation of an X-ray source (50), and an X-ray tube amplifier (52) around a patient (60) or any other object that has to be analysed. The first arm (11) bears the X-ray source (50) while the second arm (12) bears the amplifier (52). The attaching point of the two opposite arms (11,12) defines a rotation axis that usually lies in a horizontal plane. The patient (60) rests on a horizontal table (14) and the height if the table (14) is set in order that the rotation axis of the apparatus (10) be aligned as close as possible with the center of the analysed organ. This setting maintains the X-ray source (50), the organ and the X-ray tube amplifier (52) constantly in registry with each other, independently of the rotation of the arms (11,12). In use, the X-ray source (50) generates an X-ray beam aimed at the X-ray tube amplifier (52). A collimator in the X-ray source (50) is used to limit the width of the beam. The organ or organs, for instance the lungs (62) of the patient (60), are scanned by the apparatus (10) at various angular positions. The data are collected by the amplifier (52) and are sent to a computer (not shown) for reconstructing the images by digital subtraction (DSA). These images will be later analysed by the specialist. Alternatively, the amplifier (52) can be replaced by a film. Internal compensation FIG. 2 shows an example of an internal compensation of the hyper transparency of the lungs (62), due to the air therein, using a compensating device (20) according to a possible embodiment of the present invention. Other organs can also be compensated in a same manner, such as the bladder (not shown), In the case of the lungs, a pair of the three-dimensional compensating filters is used, namely a left (22) and a right filter (24). Each filter (22,24) is in registry with a corresponding lung (62) of the patient (60) and is roughly a three-dimensional scaled representation of the corresponding human lung (62), preferably of an average size. The volume of the filters (22,24) is reduced in the inverse ratio of the distance to the focal point of the X-rays and the distance of the lungs (62) of the patient (60) to the focal point. Commonly, if the filter is 50% closer than the lungs (62) of the patient (60), its volume is 50% of the size thereof. A filter of 33% of the size of the lungs (62) would be located at 1/3 of the distance. It preferably has a volume between 1/2 and 1/3 of the volume of such average human lungs, depending on their relative position with reference to the patient (60). In general, it is desirable that the image density of the compensated lungs (62) attain the density of the tissues located around them, which is that of water or muscles. Therefore, the compensation required from the filters (22,24) depends on their size. As an example, if the filters (22,24) are twice as small as the lungs (62) of the patient (60), their absorption will be about twice the one of water. One possible material is polyurethane. The penumbra created by the filters (22,24) is usually sufficient for a gradual demarcation of the shadow, but a thin coating of a slightly less absorbing material (not shown) would diminish such demarcation. As shown in FIG. 1, the pair of filters (22,24) is supported by a supporting assembly comprising a plate (32) and opposite arms (34) which allow the filters (22,24) to be held in position on the first arm (11) and in front of the X-ray source (50). In accordance with the present invention, the filters (22,24) define a plane, called the filter plane (21), which remains substantially parallel to a median plane (63) defined by the organ, such as the lungs (62) of the patient (60). In the embodiment shown in FIGS. 1 and 2, both planes are horizontal. The supporting assembly is attached to the X-ray source (50), preferably by means of a Velcro.TM. band or any other suitable means, as apparent to a person skilled in the art. In use, the filters (22,24) remain in registry with the X-ray source (50), the lungs (62) of the patient and the amplifier (52), and simultaneously the filter plane (21) remains parallel to the median plane (63), as shown in FIG. 2. This creates a shadow of radiation over the desired portion of the lungs (62) to be X-rayed in function of the angle of the arms (11,12). FIG. 2 shows examples of three different positions for the compensating device (20) with reference to the lungs (62) of the patient (60). Preferably, the filter plane (21) remains parallel to the median plane (63) of the organ under the effect of gravity. To do so, the supporting assembly may comprise an axis (31), such as a cord or a bar, on which the filters (22,24) are hung. The axis (31) is 30 transparent to X-rays to allow X-rays to be absorbed uniformly by the filters (22,24), preventing, hence, disruption of its regular pathway. An example of such material is Nylon.TM.. The axis (31) extends between two opposite arms (34) and is preferably adjustable in height. This axis (31) can also be placed near each opposite arm (34), outside the radiation beam, and made of metal. The plate (32) is held on the axis (31), preferably by means of hooks (36) which also allow a longitudinal sliding of the plate (32). The distance between the filters (22,24) and the X-ray source (50) may also be changed if proper means are provided therefor. Of course, one may choose to provide a supporting assembly with a motorised actuator (not shown) for keeping the filter plane (21) parallel to the median plane (63). External compensation The external compensating device (20) is similar to the internal compensating device, except that peripheral filters are used to eliminate the unfiltered radiation and halation around the organ or organs to analyse. As shown in FIG. 3, the compensating device (20) comprises two elongated filters (66), preferably T-shaped, that are mounted on a support (40). The support (40) is removably attached over the X-ray source (50). The filters (66) are parallel to each other and have a heavier lower portion so that they remain parallel to a horizontal plane during the motion of the arms (11) of the apparatus (10). Alternatively, it is possible to use a motorised alignment instead of gravity. The filters (66) are freely rotatable around a longitudinal axis, preferably horizontal, extending between the two adjustable arms (42) of the support (40). The positions of the pair of arms (42) may be adjusted to obtain a larger or smaller opening. The base of the support (40) and the arms (42) are made of a material transparent to X-rays. As for the filters (66), it is possible to use a plastic material or lead. Two or more materials may also be combined together. Unlike the internal compensation, it is often required that all X-rays be blocked completely to prevent saturation around the organ to analyse. A gradual demarcation is preferred around the edge of the organ to analyse since it is not possible to always perfectly align the filters (66) with reference to the organ. In FIG. 3, one can see that the distance between the filters (66) at position A is smaller than the distance between the same filters (66) at position C. Peripheral filters set around the head (64) and the face of the patient (60) further reduce the halation. Examples of peripheral filters are shown in FIGS. 4 to 8. In FIGS. 4 and 5, the patient (60) has a cervical filter (100), a neck filter (102) and a head filter (104). A full face mask filter (106), shown in FIGS. 6 and 7, may be used in addition to the other filters (100,102,104). However, a partial face mask (108) is usually preferred since it fulfils most of the needs. This mask (108) may be used with straps (110) for immobilisation. Yet, a foam pillow (112) is usually set under the head (64) of the patient (60). The peripheral filters (100,102,104,106,108) are made of a flexible and transparent material that absorbs the radiation, such as polyurethane. This material preferably has a radiation absorption similar to that of water. The filters (100,102,104,106,108) are mainly used to attenuate the demarcation between the organ and the air, thereby reducing the chances of image saturation. They also provide some immobilisation of the patient (60). FIGS. 9 and 10 schematically illustrate the two T-shaped filters (66). These filters (66) are substantially similar to the filters. Both filters (66) are aligned and parallel. The filters (66) have been provided with a flat top portion (72) and side openings (74) to follow the contours of the front, the sides and the rear portions of the head. For instance, the filters (66) in FIG. 9 are set to compensate the areas around the sides of the head of a patient, while in FIG. 10, the filters (66) are set for compensating the areas around the front and the rear of the head. FIG. 11 shows star-shaped filters (70) as another example of a compensating device (20). Each star-shaped filter (70) comprises four sections, each section comprising two opposite portions ending with a sharp wedge-shaped end to provide a gradual demarcation. One section is vertical, one section is horizontal and two sections are obliquely disposed at a 45.degree. angle with reference to the horizontal and vertical sections. The lower portion of the vertical section is heavier to maintain the proper balance by gravity, hence to maintain the proper alignment within the whole range of positions of the arms (11,12) of the apparatus (10). They may also comprise side openings similar to the side openings (74) in FIGS. 9 and 10. As an alternative embodiment, as shown in FIG. 12, the scale of the star-shaped filters (70), or any of the compensating filters, can be reduced and inserted inside the housing of the X-ray source (50). The filters (70) are then protected from the surrounding environment. Although preferred embodiments of the invention have been described in detail herein and illustrated in the accompanying drawings, it is to be understood that the invention is not limited to these precise embodiments and that various changes and modifications may be effected therein without departing from the scope or spirit of the invention. For instance, multiple other shapes of filters than those disclosed herein can be designed for achieving the same results.
	summary
	summary
048329037	claims	1. A storage arrangement for storing uncanned irradiated nuclear fuels, comprising: (A) a plurality of storage tubes having exterior surfaces bounding individually sealed interiors in which uncanned irradiated nuclear fuels are contained; (B) means including a cooling chamber in which the fuel-containing tubes are mounted, for producing a cooling flow through the chamber and over the exterior surfaces of the tubes to cool the tubes and the fuels contained therein; (C) a pluarlity of communication means, one for each tube, each communication means extending into and communicating with the interior of a respective tube; and (D) flow control means operatively connected to the plurality of communication means for controlling the flow of a storage gas into and out of individual tubes, and for individually controlling the pressure of the storage gas in individual tubes, said flow control means including (A) a plurality of storage tubes having exterior surfaces bounding individually sealed interiors in which uncanned irradiated nuclear fuels are contained; (B) means including a cooling chamber in which the fuel-containing tubes are mounted, for producing a cooling flow through the chamber and over the exterior sufaces of the tubes to cool the tubes and the fuels contained therein; (C) a plurality of communication means, one for each tube, each communication means extending into and communicating with the interior of a respective tube; and (D) flow control means operatively connected to the plurality of communication means for controlling the flow of a storage gas constituting air into and out of individual tubes, and for individually controlling the pressure of the air, in individual tubes, said flow control means including (A) a plurality of storage tubes having exterior surfaces bounding individually sealed interiors in which uncanned irradiated nuclear fuels are contained; (B) means including a cooling chamber in which the fuel-containing tubes are mounted, for producing a cooling flow through the chamber and over the exterior surfaces of the tubes to cool the tubes and the fuels contained therein; (C) a plurality of communication means, one for each tube, each communication means extending into and communicating with the interior of a respective tube; and (D) flow control means operatively connected to the plurality of communication means for controlling the flow of a storage gas into and out of individual tubes, and for individually controllng the pressure of the storage gas in individual tubes, said flow control means including (A) a plurality of storage tubes having exterior surfaces bounding individually sealed interiors in which uncanned irradiated nuclear fuels are contained; (B) means including a cooling chamber in which the fuel-containing tubes are mounted, for producing a cooling flow through the chamber and over the exterior surfaces of the tubes to cool the tubes and the fuels contained therein; (C) a plurality of communication means, one for each tube, each communication means extending into and communicating with the interior of a respective tube; and (D) flow control means operatively connected to the plurality of communciation means for controlling the flow of a storage gas into and out of individual tubes, and for individually controlling the pressure of the storage gas in individiual tubes, said flow control means including (A) a plurality of storage tubes having exterior surfaces bounding individually sealed interiors in which uncanned irradiated nuclear fuels are contained; (B) means including a cooling chamber in which the fuel-containing tubes are mounted, for producing a cooling flow through the chamber and over the exterior surfaces of the tubes to cool the tubes and the fuels contained therein; (C) a plurality of communcation means, one for each tube, each communication means extending into and communicating with the interior of a respective tube; and (D) flow control means operatively connected to the plurality of communication means for controlling the flow of a storage gas into and out of individual tubes, and for individually controlling the pressure of the storage gas in individual tubes, said flow control means including supply and exhaust manifolds, each operatively connectd to sub-groups of the communication means to isolate the tubes in communication with the sub-groups of the communication means. (A) a plurality of elongated storage tubes, each having a closed end and an opposite openable end, and having exterior surfaces bounding individually sealed interiors in which uncanned irradiated nuclear fuels are contained, each storage tube including a plug mounted in the openable end and closing the same, and means for sealing each plug in the openable end at a pair of spaced-apart sealing locations; (B) means including a cooling chamber in which the fuel-containing tubes are mounted, for producing a cooling flow through the chamber and over the exterior surfaces of the tubes to cool the tubes and the fuels contained therein; (C) a plurality of communication means, one for each tube, each communication means extending into and communicating with the interior of a respective tube, each communication means including a connecting pipe extending from the exterior of a respective tube and into the interior of the same at a location between said sealing locations, and a connecting passage extending throught the plug; and (D) flow control means operatively connected to the plurality of communication means for controlling the flow of a storage gas into and out of individual tubes, and for individually controlling the pressure of the storage gas in individual tubes. (A) a plurality of storage tubes having exterior surfaces bounding individually sealed interiors in which uncanned irradiated nuclear fuels are contained; (B) means including a cooling chamber in which the fuel-containing tubes are mounted, for producing a cooling flow through the chamber and over the exterior surfaces of the tubes to cool the tubes and the fuels contained therein, said cooling chamber including a ceiling having a plurality of support openings, one for each tube, extending through the ceiling, each tube being supported by a respective support opening so that the respective tube extends downward from the respective support opening; (C) a pluarlity of communication means, one for each tube, each communication means extending into and communicating with the interior of a respective tube; and (D) flow control means operatively connected to the plurality of communication means for controlling the flow of a storage gas into and out of individual tubes, and for individually controlling the pressure of the storage gas in individual tubes. (A) a plurality of storage tubes having exterior surfaces bounding individually sealed interiors in which uncanned irradiated nuclear fuels are contained; (B) means including a cooling chamber in which the fuel-containing tubes are mounted, for producing a cooling flow of air through the chamber and over the exterior surfaces of the tubes to cool the tubes and the fuels contained therein; (C) an inlet for admittting the air into the cooling chamber from the environment at a low elevation, and an outlet for discharging the air into the environment at a high elevation, the cooling flow of air passing from the inlet over the tubes and then out throught the outlet; (D) a plurality of communication means, one for each tube, each communication means extending into and communicating with the interior of a respective tube; and (E) flow control means operatively connected to the plurality of communciation means for controlling the flow of a storage gas into and out of individual tubes, and for individually controlling the pressure of the storage gas in individual tubes. 2. The storage arrangement as recited in claim 1, wherein the supply means includes means for measuring the flow of the storage gas being delivered by the supply manifold. 3. The storage arrangement as recited in claim 1, wherein the supply means includes means for limiting the maximum pressure of the storage gas being delivered by the supply manifold. 4. The storage arrangement as recited in claim 1, wherein the supply manifold includes a plurality of supply manifold sections, each section delivering the storage gas to more than one of the communication means. 5. A storage arrangement for storing uncanned irradiated nuclear fuels, comprising: 6. A storage arrangement for storing uncanned irradiated nuclear fuels, comprising: 7. The storage arrangement as recited in claim 6, wherein the exhaust means includes means for filtering radioactive matter in the removed storage gas. 8. The storage arrangement as recited in claim 6, wherein the exhaust means includes means for measuring the flow of the storage gas being exhausted from the exhaust manifold. 9. The storage arrangement as recited in claim 6, wherein the exhaust means includes means for monitoring the radiation and moisture levels in the removed storage gas. 10. The storage arrangement as recited in claim 6, wherein the exhaust means includes means for preventing backflow of the removed storage gas. 11. The storage arrangement as recited in claim 6, wherein the exhaust manifold includes a plurality exhaust manifold sections, each section removing the storage gas from more than one of the communication means. 12. The storage arrangement as recited in claim 6, wherein the storage gas is air, and the exhauster removes the air and produces a subatmospheric pressure in the exhausted tubes. 13. A storage arrangement for storing uncanned irradiated nuclear fuels, comprising: 14. A storage arrangement for storing uncanned irradiated nuclear fuels, comprising: 15. A storage arrangement for storing uncanned irradiated nuclear fuels, comprising: 16. The storage arrangement as recited in claim 15; and further comprising a porous filter mounted on the plug and extending across the connecting passage. 17. A storage arrangement for storing uncanned irradiated nuclear fuels, comprising: 18. A storage arrangement for storing uncanned irradiated nuclear fuels, comprising:
	summary
043448720	claims	1. A method of removing fission products from ruthenium-containing nitric acid solutions and thereafter isolating the fission products in a solid mass, the method comprising in combination the steps of: concentrating said solution by evaporation of the nitric acid therefrom to form a fission products concentrate having a solids content in the range of fifteen to twenty-five percent by exposing the solution to a partial vacuum having a pressure below fifty millimeters of mercury and by maintaining the temperature of the solution at a level sufficient to ensure evaporation of the nitric acid at that pressure whereby the evaporated nitric acid has a ruthenium content below one part per million; condensing the nitric acid vapor having a ruthenium content below one part per million; separating the nitric acid condensate from the fission products concentrate; adding vitrifiers and ammonia derivatives capable of nitrous-free nitrate decomposition to the fission products concentrate to form a fission product-vitrifier mixture with a reduced nitrate content, and heating in an oven the fission product-vitrifier mixture to vitrify same and to thereby isolate the fission products in a solid mass. 2. A method according to claim 1, which includes the steps of diluting the fission products concentrate with water and repeating said vacuum concentration step prior to said adding step. 3. A method according to claim 1, which includes the step of carrying out said concentrating step in a wiping blade evaporator. 4. A method according to claim 1, in which said ammonia derivative is urea and is added in a quantity of about 20 to 300% Mol relative to the nitrate content of said concentrate fission products. 5. A method according to claim 1, which includes the step of drying fission product-vitrifier mixture prior to said heating step. 6. A method according to claim 1, which includes the step of purging the oven with an inert gas during said heating step.
	summary
	claims	1. A support clamp assembly, comprising:a housing with a longitudinal main body attachable to a plurality of fingers of a sensing line support without welding, the fingers extending horizontally and being located in a vertical row,a longitudinal rear surface of the main body configured to contact a distal end of each of the plurality of fingers to confine jet pump sensing lines in support grooves between the fingers,at least one access hole defined by the housing, each access hole having an opening in the rear surface of the main body and a front surface of the housing, each hole configured to be located in front of a corresponding support groove when the support clamp assembly is attached to the sensing line support, an axial centerline of each access hole being about collinear with an axial centerline of a corresponding support groove. 2. The support clamp assembly of claim 1, wherein the at least one access hole is a plurality of access holes in a row, the row positioned lengthwise along the housing. 3. The support clamp assembly of claim 2, further comprising:at least one clearance hole running lengthwise through the housing, the clearance hole penetrating a top surface and a bottom surface of the housing. 4. The support clamp assembly of claim 3, further comprising:an upper plate affixed to the top surface of the clamp assembly,a lower plate affixed to the bottom surface of the clamp assembly,a bolt running through the at least one clearance hole and securing the upper and lower plates to the clamp assembly, each clearance hole having one bolt,the upper and lower plates attachable to the sensing line support through a non-welding process. 5. The support clamp assembly of claim 4, wherein the upper and lower plates each include a tongue, each tongue matable with a notch on a distal end of one of the plurality of fingers of the sensing line support to stabilize the clamp assembly on the sensing line support. 6. The support clamp assembly of claim 4, wherein the housing is configured to attach to the plurality of fingers such that the longitudinal rear surface of the main body of the housing is positioned about perpendicular to a top and bottom surface of each horizontally extending finger. 7. The support clamp assembly of claim 1, further comprising:a pad, configured to pass through the at least one access hole and into the corresponding support groove to dampen vibration of a corresponding sensing line within the support groove. 8. The support clamp assembly of claim 7, wherein the pad has a arced rear surface matable with an outer side surface of the corresponding sensing line. 9. The support clamp assembly of claim 8, wherein the pad has a recessed area on a front side of the pad. 10. The support clamp assembly of claim 9, further comprising:a jack screw in at least one of the access holes, each jack screw having a distal end configured to fit within the recessed area of the pad, the jack screw configured to be tightened to ensure that the pad fits snugly against the sensing line. 11. The support clamp assembly of claim 10, further comprising:a dowel pin on an inner surface of the recessed area of the pad,the jack screw defining a divot on a side surface of the jack screw near the distal end of the jack screw, the divot matable with the dowel pin to lock the jack screw and the pad in place within the corresponding support groove. 12. The support clamp assembly of claim 10, further comprising:serrated edges on a circumferential surface of the jack screw,a ratchet lock spring on the front surface of the housing, the ratchet lock spring engageable with the serrated edges of the jack screw to incrementally tighten the jack screw and provide anti-rotational support for the jack screw. 13. The support clamp assembly of claim 12, wherein the jack screw and the ratchet lock spring are made of a material configured to substantially maintain its physical properties in an operating nuclear reactor environment, wherein the jack screw and ratchet lock spring are made from a material that is different from a material of the housing, the jack screw and ratchet lock spring being made from Iconel X-750, a nickel-chromium-iron alloy. 14. The support clamp assembly of claim 1, wherein, the housing is made of a material configured to substantially maintain its physical properties in an operating nuclear reactor environment, the housing material being type 316 austenitic stainless steel. 15. The support clamp assembly of claim 1, further comprising:opposing, vertically extending tongues attached to the main body of the housing, the tongues configured to insert into grooves on a side surface of the fingers to attach the housing to the fingers without welding,wherein the longitudinal rear surface of the main body is about perpendicular to the horizontally extending fingers, the tongues extending about parallel to a longitudinal length of the main body. 16. A support clamp assembly, comprising:a jet pump sensing line support configured to be affixed to a jet pump diffuser,a plurality of horizontally extending fingers projecting from the jet pump sensing line support, the fingers located in a vertical row, each pair of fingers defining a support groove between the pair of fingers,jet pump sensing lines configured to be located in each support groove,a housing with a longitudinal rear surface attachable to a distal end of the plurality of fingers without welding,a plurality of access holes in a row defined by the housing, the row of access holes positioned lengthwise along the housing, each hole located in front of a support groove, an axial centerline of each hole being about collinear with an axial centerline of a corresponding support groove. 17. The support clamp assembly of claim 16, further comprising:a pad located in each support groove, each pad having an arced rear surface that mates with a corresponding sensing line. 18. The support clamp assembly of claim 17, wherein each the pad has a recessed area on a front side of the pad. 19. The support clamp assembly of claim 18, further comprising:a jack screw in each of the access holes, each jack screw having a distal end fitting within the recessed area of a corresponding pad, the jack screw configured to be tightened to ensure that the corresponding pad fits snugly against the sensing line. 20. The support clamp assembly of claim 19, further comprising:a dowel pin on an inner surface of the recessed area of each pad,each jack screw defining a divot on a side surface of the jack screw near the distal end of the jack screw, the divot mating with the dowel pin to lock the jack screw and the pad in place within the corresponding support groove. 21. The support clamp assembly of claim 19, further comprising:serrated edges on a circumferential surface of each of the jack screws,a ratchet lock spring, for each jack screw, located on the front surface of the housing, each ratchet lock spring engaging the serrated edges of a corresponding jack screw to incrementally tighten the jack screw and provide anti-rotational support for the jack screw. 22. The support clamp assembly of claim 16, further comprising:at least one clearance hole running lengthwise through the housing, the clearance hole penetrating a top surface and a bottom surface of the housing,an upper plate affixed to the top surface of the clamp assembly,a lower plate affixed to the bottom surface of the clamp assembly,a bolt running through the at least one clearance hole and securing the upper and lower plates to the clamp assembly, each clearance hole having one bolt,the upper and the lower plates attached to the fingers through a non-welding process. 23. The support clamp assembly of claim 22, wherein the upper and lower plates each include a tongue, each tongue mating with a notch on a distal end of one of the plurality of fingers of the sensing line support to stabilize the clamp assembly on the sensing line support. 24. The support clamp assembly of claim 16, further comprising:opposing, vertically extending tongues attached to the housing, the tongues configured to insert into grooves on a side surface of the fingers to attach the housing to the fingers without welding,wherein the longitudinal rear surface of the housing is about perpendicular to the horizontally extending fingers, the tongues extending about parallel to a longitudinal length of the housing.
	summary
	claims	1. A radiation source comprising:a source of radiation distributed along a given direction;a plate having an aperture formed therein, said aperture facing the source of radiation; andmeans for moving the source along the given direction when no radiation is desired on the side of the plate away from the source. 2. A radiation source according to claim 1 wherein the plate is a flat plate. 3. A radiation source according to claim 1 wherein the plate has a thickness that is greater than the smallest dimension of the aperture. 4. A radiation source according to claim 3 wherein the thickness is more than five times as large as the smallest dimension of the aperture, such that radiation exiting the aperture is collimated in the direction of the smallest dimension of the aperture. 5. A radiation source according to claim 4wherein the thickness is more than ten times as large as the smallest dimension of the aperture. 6. A radiation source according to claim 1wherein the aperture has a slit shape, such that the radiation exiting the slit forms a collimated fan beam. 7. A radiation source according to claim 1 wherein the radiation source is displaceable linearly with respect to the aperture, such that radiation from the radiation source does not pass through the aperture. 8. A radiation source comprising:a plurality of individual sources of radiation;a plate having a plurality of apertures formed therein, each said aperture facing a respective individual source of radiation; andmeans for moving the sources such that they do not face the apertures so that they are situated between the apertures when no radiation is desired on the side of the plate away from the source. 9. A radiation source according to claim 8 wherein the plate is a flat plate. 10. A radiation source according to claim 8 wherein the plate has a thickness that is greater than the smallest dimension of the apertures. 11. A radiation source according to claim 10 wherein the thickness is more than five times as large as the smallest dimension of the apertures, such that radiation exiting the aperture is collimated in the direction of the smallest dimension of the aperture. 12. A radiation source according to claim 11 wherein the thickness is more than ten times as large as the smallest dimension of the apertures. 13. A radiation source according to claim 8 wherein the apertures have a slit shape, such that the radiation exiting the slits forms collimated fan beams. 14. A radiation source according to claim 8 wherein the individual sources are situated along a line.
	claims	1. An ion therapy machine comprising:an ion source for producing an ion beam traveling along an axis;means for varying a lateral width of the ion beam perpendicular to the axis as a function of a control signal;means for varying an axial extent of the beams when steered to portions of the patient according to a control signal;means for steering a center of the ion beam to different portions of a patient according in both lateral and axial dimensions to a control signal; anda beam controller executing a stored radiation plan to communicate control signals to the means for varying a lateral width, the means for varying an axial extent and the means for steering the ion beam, to provide a radiation treatment to the patient by applying ion beams of different widths to different portions of the patient. 2. The ion therapy machine of claim 1 further including:apparatus for rotating the axis of the ion beam to different angles with respect to the patient; andwherein the beam controller further executes the stored radiation plan to communicate control signals to the means for varying an axial extent, the means for varying the lateral width and steering means to apply ion beams with higher gradients at the distal edge of the tumor than at a center of the tumor. 3. The ion therapy machine of claim 1 wherein the means for varying a lateral width is a pair of successive quadrupole magnets and the width of the ion beam is adjusted by varying a separation of the quadrupole magnets. 4. The ion therapy machine of claim 1 wherein the means for varying an axial extent is a set of selectable mechanical ridge filters. 5. The ion therapy machine of claim 1 wherein the means for varying a lateral width is a mechanical collimator. 6. The ion therapy machine of claim 1 further including a radiation planning system receiving a dose plan for a tumor within the patient and providing the radiation plan to the beam controller such that the radiation plan applies wider beams to portions of the tumor with higher gradients of the dose plan and narrower beams to portions of the tumor with lower dose gradients of the dose plan in regions removed from edges of the tumor. 7. The ion therapy machine of claim 1 further including a radiation planning system receiving a dose plan for the patient and providing the radiation plan to the beam controller to steer the beams to place Bragg peaks of the beams at distal edges of a tumor. 8. A method of radiation treatment with an ion beam traveling along an axis comprising the steps of:executing a stored radiation treatment plan to:(a) vary a lateral width of the ion beam perpendicular to the axis according to beam width data in the radiation treatment plan;(b) vary an axial extent of the ion beam along the axis according to beam extent data in the radiation treatment plan;(c) steer the axis of the ion beam to different portions of a patient according to beam location data in the radiation treatment plan;whereby ion beams of different widths and extents are applied to at least two different portions of the patient.
	abstract	A closure device for containers for transporting radioactive substances, having a first and a second component each with a comb-like portion, wherein the comb-like portions, with the closure device locked, has a bolt element passing through them. The bolt element has a head with an accommodating recess running transversally to the longitudinal axis of the bolt element, that one of the components has a through-opening which, with the bolt passing through the comb-like portions, is aligned with an accommodating recess.
	summary
	description	1. Field of the Invention The present invention relates to a shielding technique of neutrons in a reactor core, and particularly, to a neutron shield for absorbing neutrons leaking from a nuclear reactor core at an outside of the core. 2. Description of the Related Art Conventionally, there are proposed technologies utilizing powder of a neutron absorbing material such as B4C for the use of a control rod (for example, Patent Document 1: Japanese Patent Laid-Open Publication No. 10-39072). However, as a neutron shield structure disposed outside a nuclear reactor core for absorbing neutrons leaking from the core, pellets formed by baking a neutron absorbing material such as B4C are used. However, it has not known up to the present that a powdered neutron absorbing material has been used for the neutron shield. A neutron shield is disposed outside the core of a nuclear reactor in a close packed manner so as to cover an outer periphery of the nuclear reactor core and performs a function to protect a nuclear reactor vessel from fast neutrons having high energy. However, unlike the situation for control rods, such a neutron shield is regarded as an internal structure of the reactor and is neither repaired nor replaced for a predetermined period of time, thus requiring a long-term soundness. However, for such conventional technology, when a pellet-type material is used as the neutron absorbing material, thermal expansion and swelling of the pellets due to neutron irradiation causes a structural degradation of the neutron shield, which makes it difficult to maintain the soundness of the neutron shield until the time of repair or replacement. For example, there is a risk that the pellets are crushed due to thermal expansion and swelling, and hence, damages of the neutron shield or other internal structures of the reactor are caused. In addition, it is not negligible that when the neutron absorbing material is formed into pellets, a baking process of the neutron absorbing material is required, which leads to a high manufacturing cost. The present invention was conceived in consideration of the circumstances encountered in the prior art mentioned above and an object thereof is to provide a neutron shield to be disposed outside a nuclear reactor core and excellent in structural soundness. These and other objects can be achieved according to the present invention by providing, in one aspect, a neutron shield which is disposed outside a core of a nuclear reactor and adapted to absorb neutrons leaking from the core, the neutron shield comprising: a plurality of containers each of which contains a powdered neutron absorbing material and which are stacked with one another in a vertical direction; and a cladding tube which houses the containers. In this aspect, it may be desired that each of the containers contains the powdered neutron absorbing material while keeping a space left in an upper portion of an inside of the container in the stacked state thereof. It may be desired that a clearance is formed in a mating surface between the container and the cladding tube with the container being housed in the cladding tube. The container may be configured so as to enclose a filling region of the neutron absorbing material with a clearance so as not to be tightly closed. The neutron shield may further include an upper chamber which is provided in an upper portion of an interior of the cladding tube and into which gas generated by a neutron absorbing reaction of the neutron absorbing material and having leaked to the outside of the container is guided, and also include a support mechanism provided in the upper chamber for pressing an uppermost container in the stacking direction to thereby support the respective containers, the support mechanism being adapted to permit a deformation caused by thermal expansion or swelling of the container. It may be desired that each of the containers includes a shroud tube for containing the neutron absorbing material, a cover plate for closing an upper opening of the shroud tube, and a bottom plate for closing a lower opening of the shroud tube, the cover plate of the container being provided with a depression which is partly recessed in the neutron absorbing material contained in the container, and the bottom plate of the container being fitted to the depression of the cover plate. The neutron absorbing material may be composed of B4C powder. The neutron shield of the present invention of the structures mentioned above is excellent in structural soundness of the neutron absorbing material. The nature and further characteristic features will be made clearer from the following descriptions made with reference to the embodiments represented by the accompanying drawings. Preferred embodiments of the present invention will be described hereunder with reference to the accompanying drawings. Further, it is to be noted that terms “upper”, “lower”, “right”, “left” and like terms are used herein with reference to the illustration of the drawings or in an actual usable state of the nuclear reactor core. FIG. 1 illustrates a nuclear reactor provided with neutron shield according to a first embodiment of the present invention. With reference to FIG. 1, a neutron shield of the present embodiment is provided in a nuclear reactor 20. The nuclear reactor 20 is a fast reactor which utilizes high-speed neutrons generated in association with a nuclear fission without decelerating. A fast reactor is a type of nuclear reactor which can be operated continuously for ten-odd to several tens of years, for example, continuously for about 30 years without replacing the nuclear fuel, and the reactor power of which is 30 MW to hundred-odd MW (ten thousand KW to one hundred and several tens thousand KW, in electric power). As a coolant, liquid sodium is used, and a temperature used for the coolant is higher than a temperature at which liquid sodium solidifies, and a temperature of 300 to 400 degrees in the coolant flow channel, and 500 to 550 degrees at the core side are used with some margin in a nuclear reactor vessel (reactor vessel RV). In the nuclear reactor 20, a bottomed cylindrical nuclear vessel 21 is protected by being covered with a gird vessel 23 supported by a pedestal 22. The top of the nuclear reactor vessel 21 is closed with a shield plug 24 constituting an upper plug. The nuclear reactor vessel 21 contains a core 25, and the inside of the nuclear reactor vessel 21 is filled with liquid sodium as the primary coolant 26. The core 25 is loaded with a fuel assembly 27 of nuclear fuel and is configured so as to provide an entirely cylindrical shape. The core 25 is protected by being surrounded with a core barrel 28. A cylindrical partition wall 29 is provided outside the core barrel 28 with a space, and the partition wall 29 is supported along with the core 25 and the core barrel 28 on a core support plate 30. The core support plate 30 is supported on a core support which is provided to the lower portion of the circumferential wall of the nuclear reactor vessel 21. The core 25 is provided, to the lower portion thereof, with an entrance module 33, and the upper portion of a fuel assembly 27 to be loaded in the core 25 is supported by an upper support plate 34. Further, a generally sleeve-shaped (cylindrical or annular) reflector 35 is formed between the core barrel 28 and the partition wall 29. The reflector 35 is linked to a reflector driving mechanism 37 via a driving shaft 36. The reflector driving mechanism 37 is placed on the upper plug forming a shield plug 24, and the reflector 35 moves vertically around the core 25 by being driven by the reflector driving mechanism 37 thereby controlling the reactivity of the core 25. Further, a reactor shutdown rod 38 is disposed at the central portion of the core 25 so as to be loaded into or unloaded from the core through the driving by a reactor-shutdown-rod driving mechanism 39. The reactor-shutdown-rod driving mechanism 39 is also placed on the upper plug along with the reflector driving mechanism 37. The reflector driving mechanism 37 and the reactor-shutdown-rod driving mechanism 39 are housed and protected in a housing (containment) dome 40. On the other hand, the outside of the partition wall 29 forms a flow channel of a primary coolant 26 between the partition wall 29 and the nuclear reactor vessel 21, and a neutron shield 7 is arranged in the flow channel. The neutron shield 7 is placed on the core support plate 30, and the upper portion of the neutron shield 7 is supported by the upper support plate 34. The neutron shield 7 shields the neutrons which pass through or circumvent the reflector 35 and are radiated from the core 25. It is noted that the neutron shield 7 is supported only in the horizontal direction by the core support plate 30 and the upper support plate 34, and is not supported in the vertical direction so as to be freely slidable by thermal expansion. In the nuclear reactor vessel 21, an annular-shaped electromagnetic pump 45 is disposed above the neutron shield 7, and above the electromagnetic pump 45, an intermediate heat exchanger 46 is placed. The electromagnetic pump 45 and the intermediate heat exchanger 46 are formed, for example, as an integral structure or in an integral manner. The electromagnetic pump 45 circulates the primary coolant 26 in the nuclear reactor vessel 21 as shown in solid arrows, to cause it to flow from upward to downward in the coolant flow channel. Further, the primary coolant 26 and secondary coolant are guided to the tube side and the shell side of an intermediate heat exchanger 46, respectively, in a heat-exchangeable manner. The secondary coolant enters from an inlet nozzle 47 toward the intermediate heat exchanger 46 and exchanges heat with the primary coolant 26 at the intermediate heat exchanger 46 to thereby increase the temperature thereof. Thereafter, the primary coolant is fed from an outlet nozzle 48 to a vapor generator outside the nuclear reactor vessel 21, which is not shown. The secondary coolant is also made up of liquid sodium as the primary coolant is. FIG. 2 is an enlarged sectional view of an essential portion of the neutron shield 7, and FIG. 3 shows a support mechanism of a container 72 provided in the neutron shield 7, in which reference numeral 77 denotes an upper end plug of a cladding tube 74, which is inserted into a hole, not shown, formed in the upper support plate 34. As shown in FIGS. 2 and 3, the neutron shield 7 is disposed outside the nuclear reactor core for absorbing fast neutrons leaking from the core. The neutron shield 7 includes a neutron absorbing material 71, a container 72, a space 73, a cladding tube 74, a clearance 75, and a support mechanism 76. The neutron absorbing material 71 is made up of a powder of B4C (boron carbide), and is vibration-compacted in the container 72. The container 72 is provided in multiple numbers, and each container 72 contains a powder of B4C. This container 72 is configured to include a shroud tube 721, a cover plate 722, a bottom plate 723 and a joint member 724. Each container 72 is formed into a stacked structure in which the cover plate 722 of one side of the container 72 is in abutment with the bottom plate 723 of the other side of the container 72. Further, the cover plate 722 and the bottom plate 723 of the container 72 are both formed into a flat plate shape. The shroud tube 721 of the container 72 is made of austenitic stainless steel and has a thickness of about 0.2 mm and an axial length of about 1 m per one shroud tube 721. The cover plate 722 and the bottom plate 723 of the container 72 are joined with the shroud tube 721 at the joint portion 724 so as to enclose upper and lower openings of the shroud tube 721. As a joining method, a resistance welding such as a spot welding and a laser welding which has less thermal effects and thermal deformation may be used. A clearance is formed in the joint portion 724 so that the container 72 does not constitute a tightly closed structure. The space 73 is formed, without utilizing any partition structure, by adopting a structure in which the interior of the container 72 is not fully filled with neutron absorbing material 71. The cladding tube 74 is dimensioned to have a diameter of about 200 mm and a total length of about 8 m suitable for housing the container 72. Then, a clearance 75 of about 0.5 mm is provided on the mating surface between the container 72 and the cladding tube 74 with the container 72 being housed in the cladding tube 74. The cladding tube 74 is made of cold worked austenitic stainless steel or HT-9 in order to suppress a swelling phenomenon due to neutron irradiation. Further, as shown in FIG. 3, an upper chamber 741 is provided in the upper portion of the cladding tube 74, and a support mechanism 76 of the container 72 is provided in the upper chamber 741. The support mechanism 76 is configured, as shown in FIG. 3, to downwardly press the cover plate 722 of the container 72 in the uppermost layer so as to support a stacked structure of the containers 72 and to permit its displacement in the upward direction in accordance with deformation of the container 72 caused by the thermal expansion or swelling of the container 72 and other factors. The support mechanism 76 is made of a base 761 attached to the upper end plug 77 of the cladding tube 74, a guide rod 763 attached to the base 761 by a fastener 762, and a guide 764 and a spring 765 which are pressure welded to the cover plate 722 of the container 72. The neutron shield 7 of the structure mentioned above will attain the following effects. (1) The neutron shield 7 includes a plurality of containers 72 each of which contains a powdered neutron absorbing material 71 (for example, B4C) and which are stacked in the vertical direction, and a cladding tube 74 which houses the containers 72. Thus, it becomes possible to contain helium gas, which is generated by, for example, the 10B(n, α)7Li reaction inside the container 72, in a fine clearance of the powder-filled region. Since a plurality of independent containers 72 are stacked with one another, the neutron shield 7 is excellent in the mechanical stability of the containers 72, thus also being excellent in the structural soundness of the neutron absorber (neutron absorbing material) for the neutron shield disposed outside the reactor core. On the other hand, the filling density of the neutron absorbing material 71 in the container 72 is set to a predetermined value based on the design specification. Therefore, in the case of manufacturing the neutron shield 7 by vibration-compacting the neutron absorbing material 71 into the container 72 having a length similar to that of the cladding tube 74, it takes a very long time to obtain a uniform filling density in the vertical direction. Further, the neutron shield 7 is intended to be used for a long period of time without scheduled repair and replacement during the operational life of the nuclear reactor 20, there may be a decline in the balance of specific gravity due to nuclear reaction products and a change in filling density due to gravitational effects after being used for a long period of time. The neutron shield 7 of the nuclear reactor 20 has a structure in which the neutron absorbing material 71 is contained in a plurality of independent containers 72 which are stacked with one another. For this reason, it becomes easy to manufacture the neutron shield 7 which can effectively prevent a degradation of the uniform distribution of the neutron absorbing material 71 in the interior of the container 72. Furthermore, it also becomes easy to manufacture a neutron shield 7 which varies the filling density in the vertical direction. In addition, different from a pellet-type neutron absorbing material 71, a baking process may be eliminated in the manufacturing of neutron absorber, thereby reducing the manufacturing cost. (2) The container 72 contains the powdered neutron absorbing material 71 with a space 73 left in the upper portion therein. As a result, any deformation due to expansion and contraction of the container 72, which may be caused by heat dispersion and swelling of the neutron absorbing material 71 due to the neutron irradiation, can be absorbed, thereby further enhancing the effect (1) mentioned above. (3) The clearance 75 of about 0.5 mm is formed in the mating surface between the container 72 and the cladding tube 74, for example, with the container 72 being housed in the cladding tube 74. As a result, even if the deformation due to the expansion or contraction takes place in the container 72, it becomes difficult for a load to be applied to the cladding tube 74, thus also enhancing the effect (1) mentioned above. (4) The container 72 is configured so as not to tightly enclose the filling region of the neutron absorbing material 71. Therefore, helium gas generated by the 10B(n,α)7Li reaction inside the container 72 will escape to the outside of the cladding tube 721, thereby alleviating the rising of the internal pressure of the container 72. As a result, the effect (1) mentioned above can be further enhanced. (5) The neutron shield 7 includes an upper chamber 741 which is disposed in the upper portion in the cladding 74 and to which helium gas etc. generated by the 10B(n,α)7Li reaction and leaking to the outside of the container 72 is guided, and also includes a support mechanism 76 which is provided in the upper chamber 741 and which supports the containers 72 by pressing the uppermost container 72 in the stacking direction of the containers and permits deformation caused by thermal expansion or swelling of the container 72 or other factors. That is, by utilizing the upper chamber 741, which is the installation space for the support mechanism 76 to prevent the vibration of the container 72, the helium gas having escaped from the inside to the outside of the container 72 through the joint part 724 can be guided and reserved. Accordingly, by appropriately setting the volume of the upper chamber 741, it becomes possible to alleviate the rising of the internal pressure of the cladding tube 74 while maintaining the enclosure of the helium gas etc. FIG. 4 shows a second embodiment of the neutron shield relating to the present invention. The second embodiment is an example in which the configurations of the cover plate 722 and the bottom plate 723 are modified in the container 72 of the first embodiment. It is noted that the similar configurations to those of the first embodiment are given like reference symbols will be described by attaching “A” at the end of the reference symbol, and the duplicated descriptions will be simplified or omitted herein. The neutron shield 7A includes a container 72A as shown in FIG. 4. The container 72A includes a cover plate 722A and a bottom plate 723A. The cover plate 722A of the container 72A includes a depressed portion which is partly recessed so as to be buried in the neutron absorbing material 71 contained in the container 72A so as to provide a downwardly convex shape in section. On the other hand, the bottom plate 723A of the container 72A is configured to have a downwardly convex shape in section so as to fit into the depressed portion of the cover plate 722A. The neutron shield 7A of this embodiment will perform the following effects in addition to the effects (1) to (5) mentioned above of the first embodiment. (6) In the neutrons which are generated at the core 25 (see FIG. 1) and move toward the neutron shield 7A, the number of the neutrons passing through the space 73 of the container 72A is decreased, so that the neutron absorbing effect achieving the effect of (2) of the first embodiment can be prevented from lowering. FIG. 5 shows a third embodiment of the neutron shield relating to the present invention. The third embodiment is an example in which the configurations of the cover plate 722 and the bottom plate 723 are modified in the container 72 of the first embodiment. It is noted that the similar configurations to those of the first embodiment are given like reference symbols will be described by attaching “B” at the end of the reference symbol, and the duplicated descriptions will be simplified or omitted herein. The neutron shield 7B includes a container 72B as shown in FIG. 4. The container 72B includes a cover plate 722B and a bottom plate 723B. The cover plate 722B of the container 72B includes a depressed portion which is partly recessed so as to be buried in the neutron absorbing material 71 contained in the container 72B so as to provide a downwardly convex V-shape in section. On the other hand, the bottom plate 723B of the container 72B is configured to have a downwardly convex V-shape in section so as to fit into the depressed portion of the cover plate 722B. The neutron shield 7B of this embodiment will perform the following effects in addition to the effects (1) to (5) of the first embodiment and the effect (6) of the second embodiment mentioned above. (7) In forming a V-shaped portion in the cover plate 722B of the container 72B, it is not necessary to use any welding means, and it is also not necessary to perform much machining working amount, resulting in improvement in manufacturability. It is to be noted that although the neutron shield, the nuclear reactor, and the neutron absorbing method relating to the present invention have been described hereinabove with reference to the first to third embodiments, the present invention is not limited to those embodiments, particularly, in their structures, and many other changes and modifications may be made without departing from the scopes of the present invention defined in the appended claims, for example. For example, in the first embodiment, although an example of using a powder of B4C as the neutron absorbing material is shown, it may be possible to use a powder of silver alloy or powder of cadmium compound. Further, the number of stacks of the container 72 is not specifically limited.
046997504	claims	1. In a fuel assembly guide thimble inspecting system having a plurality of drag gages being insertable within guide thimbles of a fuel assembly, a drag gage handling apparatus comprising: (a) storing means for holding the plurality of drag gages in a series of storage positions being angularly displaced from one another about a common axis; (b) means for moving said drag gages from said respective storage positions about an endless path to dispose a selected one of said gages at a retrieval-and-return station; (c) loading means operable for respectively gripping and releasing said selected one drag gage at said retrieval-and-return station; (d) transporting means movable between a work station and a storage station; (e) hoisting means supporting said loading means and being supported by said transporting means, said hoisting means being operable for respectively raising and lowering said loading means and said selected one drag gage therewith away from and toward said retrieval-and-return station when said transporting means is at said storage station and operable for respectively raising and lowering said loading means and said selected one drag gage away from and toward the fuel assembly guide thimbles when said transporting means is at said work station; (f) safety means releasably mounted to said loading means for restraining said loading means in gripping relation with said selected one drag gage; and (g) means located adjacent to said retrieval-and-return station for releasing said safety means from restraining said loading means in gripping relation with said selected one drag gage only when said loading means and selected one drag gage are disposed at said retrieval-and-return station. a load cell interfaced between said loading means and said hoisting means for measuring the weight of said selected one drag gage as it is inserted within said fuel assembly guide thimbles. (a) an upright support structure; (b) a spider rotatably mounted to said support structure supporting the plurality of drag gages in a series of storage positions being angularly displaced from one another about said support structure; (c) drive means for rotating said spider relative to said support structure and thereby move said drag gages supported on said spider from said respective storage positions about an endless path to dispose a selected one of said gages at a retrieval-and-return station; (d) a gripper mechanism operable for respectively gripping and releasing said drag gage at said retrieval-and-return station; (e) an elongated track mounted to and extending outwardly from said support structure above said spider; (f) a trolley movable along said track between a position remote from said support structure and overlying a work station and a position adjacent to said support structure and overlying said retrieval-and-return station; (g) a hoist supporting said gripper mechanism and being supported by said trolley, said hoist being operable for respectively raising and lowering said gripper mechanism and said selected one drag gage therewith away from and toward said retrieval-and-return station when said trolley is at said position adjacent said support structure and operable for respectively raising and lowering said gripper mechanism and said selected one drag gage therewith away from and toward the fuel assembly guide thimbles when said trolley is at said remote position; (h) a safety sleeve normally disposed about said gripper mechanism for restraining said gripper mechanism in gripping relation with said selected one drag gage; and (i) an arm structure mounted to said support structure so as to engage and displace said safety sleeve from about said gripper mechanism so as to permit release of said gripper mechanism from its gripping relation with said selected one drag gage when said gripper mechanism and said selected one drag gage are disposed at said retrieval-and-return station. a load cell interfaced between said gripper mechanism and said hoist for measuring the weight of said selected one drag gage as it is inserted within said fuel assembly guide thimbles. a set of angularly displaced gripping fingers mounted for pivotal movement radially relative to a common axis between an inwardly-displaced contracted position for gripping said selected one drag gage and an outwardly-displaced expanded position for releasing said gage; and means operable for moving said gripping fingers between said respective positions. an actuator being linked to said set of fingers such that when actuated said actuator causes the same to move to their contracted gripping position; a sensor for sensing when said fingers are disposed adjacent said selected one drag gage for gripping the same; and a resiliently yieldable member connected to said fingers for returning the same to their expanded releasing position whenever said actuator is deactuated. (a) storing means for holding the plurality of drag gages in a series of storage positions being angularly displaced from one another about a common axis; and (b) means for moving said drag gages from said respective storage positions about an endless path to dispose a selected one of said gages at a retrieval-and-return station; (c) said storing means including a plurality of radially-projecting arms each having a cradle attached to an outer end thereof for receiving and supporting one of said drag gages. (a) an upright support structure; (b) a spider rotatably mounted to said support structure for supporting the plurality of drag gages in a series of storage positions being angularly displaced from one another about said support structure; and (c) drive means for rotating said spider relative to said support structure and thereby move said drag gages supported on said spider from said respective positions about an endless path to dispose a selected one of said gages at a retrieval-and-return station; (d) said spider including a plurality of radially-projecting arms each having a cradle attached to an outer end thereof for receiving and supporting one of said drag gages. a motor mounted to said support structure adjacent to said spider for providing a source of rotary motion; and interengaging means coupling said motor to said spider for transmitting rotary motion of said motor to said spider for rotatably driving the same relative to said support structure. (a) a gripper mechanism actuatable between contracted and expanded positions for respectively gripping and releasing a selected one of said drag gages; (b) safety means for restraining said gripper mechanism at its contracted position in gripping relation with said selected one drag gage; and (c) means for releasing said safety means from restraining said gripper mechanism at said contracted position in gripping relation with said selected one drag gage only when disposed at a retrieval-and-return station for facilitating actuation of said gripper mechanism to said expanded position in releasing relation with said selected one gage and thereby facilitating respective fetching and storing thereof; (d) said safety means being a safety sleeve normally disposed about said gripper mechanism for restraining said gripper mechanism in gripping relation with said selected one drag gage, said sleeve being displacable from about said gripper mechanism by said releasing means. a set of angularly displaced gripping fingers mounted for pivotal movement radially relative to a common axis between an inwardly-displaced contracted position for gripping said selected one drag gage and an outwardly-displaced expanded position for releasing said gage; and means operable for moving said gripping fingers between said respective positions. an upright support structure adjacent said retrieval-and-return stations; and an arm structure mounted to said support structure so as to engage and displace said safety sleeve from about said gripper mechanism so as to permit release of said gripper mechansim from its gripping relation with said selected one drag gage when disposed at said retrieval-and-return station. 2. The handling apparatus as recited in claim 1, further comprising: 3. In a fuel assembly guide thimble inspecting system having a plurality of drag gages being insertable within guide thimbles of a fuel assembly, a drag gage handling apparatus comprising: 4. The handling apparatus as recited in claim 3, further comprising: 5. The handling apparatus as recited in claim 3, wherein said gripper mechanism includes: 6. The handling apparatus as recited in claim 5, wherein said operable means includes: 7. In a fuel assembly guide thimble inspecting system having a plurality of drag gages being insertable within guide thimbles of a fuel assembly, a drag gage dispenser comprising: 8. In a fuel assembly guide thimble inspecting system having a plurality of drag gages being insertable within guide thimbles of a fuel assembly, a drag gage dispenser comprising: 9. The dispenser as recited in claim 8, wherein said drive means includes: 10. In a fuel assembly guide thimble inspecting system having a plurality of drag gages being insertable within guide thimbles of a fuel assembly, a safety mechanism comprising: 11. The safety mechanism as recited in claim 10, wherein said gripper mechanism includes: 12. The safety mechanism as recited in claim 10, wherein said releasing means includes:
	description	This invention pertains in general to nuclear reactor containment systems, and, more particularly, to nuclear reactor containment ventilation filtration systems for safely reducing pressure buildup within a containment without the release of harmful contaminants. In many jurisdictions, nuclear power plants must be designed to ensure that even in the event of accidents, a mechanism will be provided to prevent or minimize the escape of radioactive material and noble gases. To guard against radioactive releases, the reactor system is typically housed within a primary containment structure that is constructed from steel and reinforced concrete. The primary containment vessel is designed to be capable of withstanding large pressures which may result from various postulated accident scenarios. However, if a scenario is postulated to be sufficiently severe, the containment vessel itself could fail from gradually increasing pressure. Although the likelihood of such an event is considered very small, the health risks associated with exposing the surrounding population to the radioactive releases of such an event has led many to believe that a mechanism should be provided to vent the containment in a controlled manner and filter the gases to minimize the release of radioactivity. That is, it is desirable to both provide a pressure release device for the containment vessel and a mechanism for filtering any gases that may be released by the containment before they are released into the atmosphere. The nuclear accidents at Chernobyl, Ukraine in 1986 and more recently Fukushima Dai-ichi in Japan in 2011, clearly show the consequences of a release of fission products with long decay times. As a result of these accidents, the governments of many countries have decided that nuclear power plants must install filtered containment ventilation systems to protect people and the surrounding land from damage due to radioactive contamination. In the past, a number of filtration systems have been proposed, such as the one described in U.S. Pat. No. 4,610,840, issued to Leach and assigned to the Assignee of this invention. Leach discloses a fission product scrubbing system for a nuclear reactor. Specifically, a second compartment in fluid communication with the containment is partially filled with water. In the event of a large pressure increase, a ruptured disc disposed within a vent pipe emanating from the secondary compartment bursts to relieve pressure. When the rupture disc blows, radioactive gases and vapors from the containment pass through the water filled secondary compartment and are then released through the now open vent pipe. As the hot containment gases and vapors pass through the water stored within the enclosed secondary compartment, a large portion of the fission products will be scrubbed from the containment gases. While such a system can be effective, there is still room for improvement for reducing the size and increasing the effectiveness of such a system to minimize any exposure that such a release may potentially cause. U.S. Pat. No. 9,502,144, assigned to the Assignee of this invention, describes a filter system with such an improvement but still requires a significant investment to implement. Accordingly, a filtering system is desired that will effectively filter the release of the containment atmosphere to the external environment that requires a substantially reduced investment to implement. This invention provides a filtered venting system for a nuclear power generating facility having a containment for housing a nuclear reactor and for confining radiation leaked from the nuclear reactor. The containment has a ventilation outlet for providing a controlled release to the environment surrounding the containment, for an atmospheric pressure buildup within the containment in the event the pressure of an atmospheric effluent within the containment is built up to a level that exceeded a preselected value. The nuclear power generating facility also has, outside the containment, an associated spent fuel storage water pool, including a filter system for filtering contaminants released from or on route to the ventilation outlet. The filter system includes a dedicated piping system that is connected between an interior of the containment or the ventilation outlet and the spent fuel storage water pool, for fluidly communicating any atmospheric effluent to be released from inside of the containment through the spent fuel storage water pool. The filter system also includes one or more valves connected to the dedicated piping system for controlling the release of the atmospheric effluent to be released; and a chemical injection system configured to release a chemical into the spent fuel storage water pool to facilitate a reaction with the atmospheric effluent to be released to substantially neuter any deleterious environmental impact of the atmospheric effluent to be released. A control system is connected to one or more of the chemical injection systems and/or the one or more of the valves and is configured to control the release of the chemical and/or the release of the atmospheric effluent. In one embodiment the dedicated piping system includes a check valve configured to prevent spent fuel pool water from being drawn into the containment. Preferably, the control system includes a manually actuated, remotely operated valve(s) in the dedicated piping system, the valve(s) being configured to isolate the dedicated vent piping under normal operating conditions, unless activated. Preferably, this configuration of valve(s) for isolation will be located outside of the containment building and will comply with the regulated requirements for isolation of piping that penetrates a containment building. The manually actuated, remote operated valve(s) is configured in the dedicated piping system to be in parallel with a passively operated valve structured to release the atmospheric effluent to the spent fuel water pool if a pressure is sensed within the containment in excess of a given pressure. In another embodiment the dedicated piping system has an outlet in a lower portion of the spent fuel water pool that releases the atmospheric effluent to the spent fuel water pool through a sparger. Preferably, the chemical injection system releases the chemical into the spent fuel water pool through a chemical injection header that is supported just above the sparger within the spent fuel storage water pool. Desirably, the chemical injection header and the sparger are supported within the spent fuel storage water pool at an elevation below the area that is used to transfer fuel into and out of the spent fuel storage water pool and, preferably, as low as possible to maximize transit time of the released effluent through the water pool. The invention also contemplates a method for operating such a filter system that includes the step of sensing a pressure buildup within the nuclear containment. The method routs a portion of the atmospheric effluent through the spent fuel pool when the pressure buildup within the containment reaches a preselected value and releases a chemical into the spent fuel pool to facilitate a reaction with the effluent to be released to substantially neuter any deleterious environmental impact of the effluent to be released. Preferably, the routing step and the releasing step are performed at the same time and the atmospheric effluent is introduced into water in the spent fuel pool through a sparger supported near or at a bottom of the spent fuel pool. Desirably the releasing step releases the chemical into water within the spent fuel pool at an elevation near the bottom of the spent fuel pool just above the sparger. This invention involves an application specific design of piping, valves, control logic and a chemical injection system to effectively employ the concepts of a wet filtered vent design, such as the one described in U.S. Pat. No. 9,502,144, without the addition of a wet filter vent filtration tank. One embodiment of this invention is illustrated in FIG. 1, which shows a schematic representation of a portion of a nuclear containment and adjacent spent fuel pool. This invention uses ventilation piping 10 that directs a pressure relief discharge from the containment vessel 12 into the planes existing spent fuel pool 14 through an engineered sparger design (or existing spent fuel pool cooling system sparger) 16. Isolation of the ventilation piping is achieved via conventional, remotely operated valve(s) 18, controlled to open by manual actuation by the plant operator. An alternate bypass system, with passive pressure relief valve 20, is available in the event of an operator error or mechanical failure of the isolation valve(s) 18. The bypass system automatically opens the valve 20, which is a passive pressure relief device, if a preselected pressure is sensed in the containment. The contaminated aerosol release will be filtered via the spent fuel pool inventory, which will be treated with conventional wet filtration chemistry control via a passive chemical injection system 22 for gas (e.g., iodine, cesium, xenon) and fission product particulates removal. The chemicals will be released into the pool inventory simultaneous with the ventilation release to the pool. (i.e., opening of the ventilation isolation valves 18 or 20) via a controlled opening of the chemical injection system isolation valve 24. The chemicals will be injected directly above the sparger outlets 16 via a chemical injection header 26. Preferably, the chemical injection header and the sparger are supported in the spent fuel pool at an elevation, preferably, as low as possible in the pool and below the operating level necessary for fuel transfer into and out of the pool. The chemical injection header 26 is, preferably, positioned just above and over the sparger 16. The consequential fission product decay heat energy released to the pool will be removed by the current spent fuel pool cooling system. Aerosol release from the spent fuel pool surface will be vented from the spent fuel building via normal or special supplemental, if necessary, spent fuel pool ventilation systems. Liquid swell in the spent fuel pool will not be sufficient to displace excessive pool inventory such that acceptable spent fuel bundle submergence remains for shielding the spent fuel following closure of the vent isolation valves. Check valve(s) (passive dampers) 28 in the ventilation piping will prevent draw of pool inventory into containment during any containment vessel vacuum; similarly, a passive vacuum breaker 32 will prevent the containment vessel from exceeding a maximum vacuum limit. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. An example could be to avoid installation of a unique containment vessel penetration, an alternate embodiment could incorporate the device into the existing fuel transfer tube 30. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
	abstract	A diode assembly for producing a pulsed fusion event in a z-pinch driver. The diode assembly includes an inner core formed of a fusionable fuel source material including a lithium compound formed of one or more lithium isotopes and one or more hydrogen isotopes. A lithium metal outer sheath is integrally formed around the inner core by decomposing a surface of the fusionable fuel source material.
	summary
	summary
046997555	abstract	Ultrafiltration circuit for the primary cooling fluid of a pressurized-water nuclear reactor, comprising a first loop (27) taken off from the discharge branch (10) of the volumetric and chemical monitoring circuit (8) and a second loop (28) taken off from the charge branch (11) of the circuit (8). Each of the loops (27, 28) incorporates an ultrafilter (30, 50). The concentrate from the ultrafilter (30) located on the first loop (27) is cooled, depressurized and conveyed into the chemical monitoring circuit (8) via a pipe (38) downstream of the main cooling and depressurizing device (12, 13) of the circuit (8). The concentrate from the ultrafilter (50) of the loop (28) is conveyed into the chemical monitoring circuit (98) via a pipe (58) upstream of the device (12, 13).
062367005	claims	1. A downcommer coupling apparatus for coupling a downcommer pipe of a core spray line to a core shroud in a nuclear reactor, the downcommer pipe comprising at least one dowel bolt opening said apparatus comprising: a wedge housing configured to engage a downcommer pipe, said wedge housing comprising at least one dowel bolt opening; a wedge flange engaging said wedge housing; a wedge having a plurality of segments, said wedge extending into said wedge housing; and at least one dowel bolt, each said dowel bolt extending through a dowel bolt opening in said housing, each said dowel bolt configured to extend through a dowel bolt opening in the downcommer pipe. a wedge housing comprising a frusto-conical portion and a flange located at a large diameter end of said frusto-conical portion, said wedge housing further comprising at least one threaded dowel bolt opening, each said dowel bolt opening configured to substantially align with a dowel bolt opening in a downcommer pipe; a wedge comprising a plurality of segments sized to extend into said wedge housing; a wedge flange; and at least one dowel bolt extending through and threadedly engaging said wedge housing dowel bolt opening and configured to extend through the at least one dowel bolt opening in the downcommer pipe. 2. A downcommer coupling apparatus in accordance with claim 1 wherein said wedge flange and wedge housing each comprise at least one wedge flange bolt opening, and wherein said apparatus further comprises at least one wedge flange bolt configured to extend through said wedge flange bolt openings when aligned. 3. A downcommer coupling apparatus in accordance with claim 1 further comprising a cylindrical pipe and an elbow, wherein said pipe first end coupled to said wedge housing and said elbow coupled to said pipe second end. 4. A downcommer coupling apparatus in accordance with claim 3 further comprising a lower flange coupled to said elbow second end, said elbow second end configured to couple to the core shroud. 5. A downcommer coupling apparatus in accordance with claim 4 wherein said lower flange comprises at least one lower flange bolt opening, and wherein said downcommer coupling apparatus further comprises at least one lower flange bolt for extending through said lower flange bolt opening. 6. A downcommer coupling apparatus in accordance with claim 5 further comprising a nutbar having at least one nut for engaging said lower flange bolt. 7. A downcommer coupling apparatus in accordance with claim 5 further comprising a clamp for engaging said lower flange bolt. 8. A downcommer coupling apparatus in accordance with claim 1 further comprising a pipe seal engaging said wedge housing, said pipe seal configured to engage the downcommer pipe. 9. A downcommer coupling apparatus in accordance with claim 8 wherein said pipe seal is a flexible double diaphragm type seal. 10. A downcommer coupling apparatus in accordance with claim 3 further comprising a shroud seal engaging said lower flange, said shroud seal configured to engage a core shroud. 11. A downcommer coupling apparatus for coupling a downcommer pipe of a core spray line to a core shroud in a nuclear reactor, the downcommer pipe including at least one dowel bolt opening, said apparatus comprising: 12. An apparatus in accordance with claim 11 wherein said wedge flange and said wedge housing flange each comprise at least one flange bolt opening, said apparatus further comprising at least one flange bolt extending through said at least one opening in said wedge flange and said wedge housing flange. 13. An apparatus in accordance with claim 11 further comprising a cylindrical pipe having a first end and a second end, and an elbow having a first end and a second end, said pipe first end coupled to said wedge housing and said pipe second end coupled to said elbow first end. 14. An apparatus in accordance with claim 11 further comprising a lower flange coupled to said second end of said elbow. 15. An apparatus in accordance with claim 14 wherein said lower flange comprises at least one bolt opening, and said apparatus further comprising at least one lower flange bolt extending through said at least one bolt opening in said lower flange. 16. An apparatus in accordance with claim 15 further comprising a nutbar comprising at least one nut theadedly engaging said at least one lower flange bolt. 17. An apparatus in accordance with claim 15 further comprising a clamp engaging said at least one lower flange bolt. 18. An apparatus in accordance with claim 11 further comprising a pipe seal engaging said wedge housing and configured to engage the downcommer pipe. 19. An apparatus in accordance with claim 18 wherein said pipe seal comprises a flexible double diaphragm type pipe seal. 20. An apparatus in accordance with claim 14 further comprising a shroud seal engaging said lower flange and configured to engage a shroud.
041994043	claims	1. In a breeder-reactor fuel pellet of the type formed by compressing and sintering a blend of fissile and fertile components and loaded into fuel rods of a thin metal cladding, the improvement wherein: a. at least 90 percent of the fissile component content of the fuel pellet is contained in a powder consisting essentially of a blend of fissile and fertile components having a fissile-component percentage between 20 and 40 percent, the powder having a stable microstructure and a density less than about 85 percent of theoretical; and b. at least half of the fertile component content of the fuel pellet is contained in particles greater than about 50 microns in diameter, the average of the individual densities of the particles being greater than about 95 percent of theoretical. 2. A pellet as recited in claim 1, wherein the ratio of fertile atoms to fissile atoms is between 4:1 and 9:1. 3. A pellet as recited in claim 2, wherein the powder consists essentially of a blend of fertile and fissile components having a fissile-component percentage between 25 percent and 35 percent. 4. A pellet as recited in claim 3, wherein the powder consists essentially of a fertile-fissile blend selected from the group consisting of uranium-plutonium carbide, uranium-plutonium nitride, uranium-plutonium silicide, and uranium-plutonium oxide. 5. A pellet as recited in claim 3, wherein the particles consist essentially of a fertile material selected from the group consisting of uranium carbide, uranium nitride, uranium silicide, and uranium oxide. 6. A pellet as recited in claim 4, wherein the powder consists essentially of uranium-plutonium carbide and the particles consist essentially of uranium carbide.
	abstract	A control circuit administers a radiation treatment plan that specifies a planned total radiation dose for a radiation treatment session for a given patient by modulating a radiation beam with at least one high-resolution aperture that is formed using one of a plurality of linearly-sequential high-resolution aperture possibilities. By one approach the foregoing comprises modulating the radiation beam using at least substantially only high-resolution apertures that are formed using a plurality of the linearly-sequential high-resolution aperture possibilities. In some cases the foregoing can comprise administering the radiation treatment plan using at least two separate radiation exposures for only a single treatment field, in which case, by one approach, each of the separate radiation exposures for the single treatment field can comprise modulating the radiation beam using at least substantially only high-resolution apertures.
	claims	1. A variable shaped electron beam lithography system comprising a first, second and third aperture for forming a single electron beam in each of the rectangular opening portion that are provided, and for drawing a figure pattern using a single electron beam shot formed by passing the beam through the first, second and third aperture in sequence, said system comprising:a rotary drive mechanism provided in each of the first and second aperture, for rotationally driving each of the apertures around an optical axis up to an arbitrary angle from 0 to 360°; anda variable slit width mechanism for varying an opening slit width of the rectangular opening portion provided in the third aperture. 2. The variable shaped electron beam lithography system according to claim 1, whereinwhen drawing a figure pattern including two opposite, oblique sides each of which forms an arbitrary angle with a coordinate axis in an XY drawing rectangular coordinate system, on a drawn target substrate,a position at which each side of the rectangular opening portion of each aperture becomes parallel or vertical relative to the coordinate axis in the XY drawing rectangular coordinate system is defined as a rotary reference for each aperture,the first aperture is rotated by the rotary drive mechanism from the rotary reference by an angle coincident with an angle of one of the two oblique sides, and the second aperture is rotated by the rotary drive mechanism from the rotary reference by an angle coincident with an angle of the other of the two oblique sides; anda single electron beam is formed into a shape of the figure pattern using the rectangular opening portions of the first and second aperture that have been rotated, and the rectangular opening portion of the third aperture, and is projected onto the drawn target substrate. 3. The variable shaped electron beam lithography system according to claim 1, whereinwhen drawing, on the drawn target substrate, a figure pattern which is a quadrangle and composed of two opposite, oblique sides each of which makes an arbitrary angle relative to the coordinate axis in the XY drawing rectangular coordinate system and two opposite, parallel sides parallel or vertical to the coordinate axis in the XY drawing rectangular coordinate system,a position at which each side of the rectangular opening portion of each aperture becomes parallel or vertical to the coordinate axis in the XY drawing rectangular coordinate system is defined as a rotary reference for each aperture,the first aperture is rotated by the rotary drive mechanism from the rotary reference by an angle coincident with an angle of one of the two oblique sides, and the second aperture is rotated by the rotary drive mechanism from the rotary reference by an angle coincident with an angle of the other of the two oblique sides;the opening slit width of the rectangular opening portion of the third aperture is changed by the variable slit width mechanism to a dimension obtained by multiplying a distance between the two parallel sides by the inverse of the reduction rate of a reduction lens system for reducing and projecting the single electron beam, formed by passing the beam through the first, second and third aperture in sequence, on the drawn target substrate; andthe single electron beam is formed into a shape of the figure pattern using the rectangular opening portions of the first and second aperture that has been rotated, and the rectangular opening portion of the third aperture, subsequently is reduced to be the same size as the figure pattern by the reduction lens system, and is projected onto the drawn target substrate. 4. The variable shaped electron beam lithography system according to claim 3, further comprising:a rotary drive mechanism for rotationally driving the third aperture around an optical axis up to an arbitrary angle from 0 to 360°. 5. The variable shaped electron beam lithography system according to claim 4, whereinin drawing a rotated figure pattern formed by rotating the figure pattern of the quadrangle around the center of the figure thereof by an arbitrary angle on the drawn target substrate,the single electron beam is formed into a shape of the rotated figure pattern, by rotating each aperture by each rotary drive mechanism from the rotary reference by a value of the sum of the rotation angle of each aperture when the figure pattern of the quadrangle is drawn, and when the angle is the same as the rotation angle of the rotated figure pattern. 6. A method for drawing a figure pattern by variable shaped electron beam lithography system, wherein a figure pattern to be drawn on a drawn target substrate is divided into a plurality of figure portions and for each of the divided figure portions, a single electron beam is formed into the shape of the divided figure portion to draw each of the figure portions, said method comprising the steps of:varying an opening slit width of a rectangular opening portion provided in a third aperture;forming the single electron beam into the same shape as the oblique line figure portion without dividing the oblique line figure portion; anddrawing the figure pattern on the drawn target substrate, whereinthe figure pattern includes the oblique line figure portion including two opposite, oblique lines each of which makes an arbitrary, different angle with a coordinate axis in an XY drawing rectangular coordinate system. 7. A method for drawing a figure pattern by variable shaped electron beam lithography system, wherein a figure pattern to be drawn on a drawn target substrate is divided into a plurality of figure portions and for each of the divided figure portions, a single electron beam is formed into the shape of the figure portion to draw each of the figure portions, said method comprising the steps of:varying an opening slit width of a rectangular opening portion provided in a third aperture;approximating a drawing region between an outer peripheral line and an inner peripheral line of a ring pattern by a plurality of the same trapezoidal figures equally divided;forming the single electron beam for each of the divided trapezoidal figures into the same shape of the trapezoidal figure; anddrawing the ring pattern on the drawn target substrate in turn. 8. The method for drawing a figure pattern by variable shaped electron beam lithography system according to claim 6 or 7, whereinthe drawn target substrate is a dry plate. 9. The method for drawing a figure pattern by variable shaped electron beam lithography system according to claim 6 or 7, whereinthe drawn target substrate is a semiconductor wafer. 10. The method for drawing a figure pattern by variable shaped electron beam lithography system according to claim 6, further comprising:rotationally driving a first and a second apertures around an optical axis up to an arbitrary angle from 0 to 360°. 11. The method for drawing a figure pattern by variable shaped electron beam lithography system according to claim 7, further comprising:rotationally driving a first and a second apertures around an optical axis up to an arbitrary angle from 0 to 360°.
053274709	abstract	A cross point spacer is provided with subchannel vertical tubes at the subchannel volumes, these tubes being supplied with an inner flow free cylindrical volume and an peripheral steam separator in the form of a surface which imparts a spiral flow pattern to steam/water mixture flowing peripherally upwardly within the subchannel tube. According to a first and preferred embodiment, the subchannel tube of the cross point spacer is provided with rifling. According to a second embodiment, twisted vanes are installed and fastened to the interior of the subchannel tube and drilled away at the central portion of the twisted vanes to leave strips of metal defining rifling like paths on the subchannel tube interior. In either case, the main and central portion of the subchannel volume through the spacer is provided with an unobstructed central portion for upward steam flow. At the same time, the peripheral portion of the subchannel tube imparts a steam separator centrifugal upward spiral flow to water particles within the subchannel tube. Upon exit of the water and generated steam from the vertically disposed subchannel tube, a separated steam flow overlying the subchannel tube occurs. At the same time, entrained water is centrifugally classified outwardly from the steam flow to the surrounding steam generating fuel rods for the generation of steam.
	claims	1. An extreme ultra violet light source apparatus for generating extreme ultra violet light by applying a laser beam to a target material to turn the target material into plasma, said apparatus comprising:an extreme ultra violet light generating chamber in which extreme ultra violet light is generated;a target material supply unit for injecting a target material into said extreme ultra violet light generating chamber when the extreme ultra violet light is generated;a driver laser for emitting a laser beam;a window provided to said extreme ultra violet light generating chamber, for passing the laser beam into said extreme ultra violet light generating chamber;laser beam focusing optics including at least one optical element, said laser beam focusing optics focusing the laser beam emitted from said driver laser onto a trajectory of the target material injected into said extreme ultra violet light generating chamber to generate plasma;extreme ultra violet light collecting optics for collecting and outputting extreme ultra violet light radiated from said plasma;a temperature sensor for detecting a temperature of said window; anda processing unit for determining deterioration of said window based on the temperature of said window detected by said temperature sensor when extreme ultra violet light is generated. 2. The extreme ultra violet light source apparatus according to claim 1, wherein said processing unit outputs a control signal for stopping emission of the laser beam to said driver laser, when determining that said window is deteriorated. 3. The extreme ultra violet light source apparatus according to claim 1, further comprising:a laser beam detector provided outside said extreme ultra violet light generating chamber, for detecting intensity of the laser beam, which has been dispersed without being applied to the target material after being focused by said laser beam focusing optics and output from said extreme ultra violet light generating chamber, when extreme ultra violet light is not generated,wherein said processing unit determines deterioration of said window and/or said at least one optical element based on the intensity of the laser beam detected by said laser beam detector when extreme ultra violet light is not generated. 4. The extreme ultra violet light source apparatus according to claim 3, wherein said processing unit outputs a control signal for stopping emission of the laser beam to said driver laser when determining that said window is deteriorated. 5. The extreme ultra violet light source apparatus according to claim 3, further comprising:a shielding means for shielding materials and electromagnetic waves generated in said extreme ultra violet light generating chamber to protect said laser beam detector when extreme ultra violet light is generated. 6. The extreme ultra violet light source apparatus according to claim 1, further comprising:laser beam re-focusing optics provided outside said extreme ultra violet light generating chamber, for focusing again the laser beam, which has been dispersed without being applied to the target material after being focused by said laser beam focusing optics and output from said extreme ultra violet light generating chamber, when extreme ultra violet light is not generated; andan area sensor for detecting an image of the laser beam focused by the laser beam re-focusing optics,wherein said processing unit determines deterioration and/or distortion of said window and/or said at least one optical element, and/or determines whether a focus of the laser beam focused by said laser beam focusing optics is deviated from a position where the plasma is to be generated, based on the image of the laser beam detected by said area sensor when extreme ultra violet light is not generated. 7. The extreme ultra violet light source apparatus according to claim 6, further comprising:a visible fluorescent screen for converting the laser beam focused by the laser beam re-focusing optics to visible light; anda visible light focusing system for focusing the visible light converted by said visible fluorescent screen onto a light-receiving surface of said area sensor. 8. The extreme ultra violet light source apparatus according to claim 6, further comprising:a shielding means for shielding materials and electromagnetic waves generated in said extreme ultra violet light generating chamber to protect said area sensor when extreme ultra violet light is generated. 9. An extreme ultra violet light source apparatus for generating extreme ultra violet light by applying a laser beam to a target material to turn the target material into plasma, said apparatus comprising:an extreme ultra violet light generating chamber in which extreme ultra violet light is generated;a target material supply unit for injecting a target material into said extreme ultra violet light generating chamber when extreme ultra violet light is generated;a driver laser for emitting a laser beam;a window provided to said extreme ultra violet light generating chamber, for passing the laser beam into said extreme ultra violet light generating chamber;laser beam focusing optics including an optical element provided in said extreme ultra violet light generating chamber, said laser beam focusing optics focusing the laser beam emitted from said driver laser onto a trajectory of the target material injected into said extreme ultra violet light generating chamber to generate plasma;extreme ultra violet light collecting optics for collecting and outputting extreme ultra violet light radiated from said plasma;a plurality of temperature sensors, each provided to respective one of said window and said optical element, for detecting temperatures thereof; anda processing unit for determining deterioration of said window and said optical element based on the temperatures of said window and said optical element detected by said plurality of temperature sensors when extreme ultra violet light is generated. 10. The extreme ultra violet light source apparatus according to claim 9, wherein said driver laser includes a pre-pulse laser and a main pulse laser, and said window and said laser beam focusing optics are provided for each of said pre-pulse laser and said main pulse laser. 11. The extreme ultra violet light source apparatus according to claim 9, wherein said processing unit outputs a control signal for stopping emission of the laser beam to said driver laser when determining that said window is deteriorated. 12. The extreme ultra violet light source apparatus according to claim 9, further comprising:a first energy detector for detecting energy of the laser beam, which has been dispersed without being applied to the target material after being focused by said laser beam focusing optics, when extreme ultra violet light is not generated,wherein said processing unit determines deterioration of said window and said optical element based on the energy of the laser beam detected by said first energy detector when extreme ultra violet light is not generated. 13. The extreme ultra violet light source apparatus according to claim 12, further comprising:a second energy detector provided outside said extreme ultra violet light generating chamber, for detecting energy of the laser beam before passing through said window,wherein said processing unit determines deterioration of said window and said optical element based on the energy of the laser beam detected by said first energy detector and the energy of the laser beam detected by said second energy detector when extreme ultra violet light is not generated. 14. The extreme ultra violet light source apparatus according to claim 12, wherein said first energy detector is provided outside said extreme ultra violet light generating chamber, and a concave mirror for guiding the laser beam, which has been dispersed without being applied to the target material, to said first energy detector is provided inside said extreme ultra violet light generating chamber. 15. The extreme ultra violet light source apparatus according to claim 14, further comprising:a protection housing provided in said extreme ultra violet light generating chamber,wherein said concave mirror is stored capable of being taken out in said protection housing, and is inserted into an optical path of the laser beam at measurement of said first energy detector. 16. The extreme ultra violet light source apparatus according to claim 12, wherein the target material supply unit does not inject the target material while said driver laser emits the laser beam, when said extreme ultra violet light is not generated. 17. The extreme ultra violet light source apparatus according to claim 12, wherein timing of injecting the target material controlled by the target material supply unit and timing of emitting the laser beam controlled by said driver laser are adjusted so as not to apply the laser beam to the target material, when said extreme ultra violet light is not generated. 18. The extreme ultra violet light source apparatus according to claim 12, wherein the trajectory of the target material controlled by the target material supply unit and an optical path of the laser beam controlled by said driver laser are adjusted so as not to apply the laser beam to the target material, when said extreme ultra violet light is not generated. 19. The extreme ultra violet light source apparatus according to claim 9, further comprising:a shielding means for shielding materials and electromagnetic waves generated in said extreme ultra violet light generating chamber to protect said window and said optical element provided in said extreme ultra violet light generating chamber when extreme ultra violet light is generated. 20. An extreme ultra violet light source apparatus for generating extreme ultra violet light by applying a laser beam to a target material to turn the target material into plasma, said apparatus comprising:an extreme ultra violet light generating chamber in which extreme ultra violet light is generated;a target material supply unit for injecting a target material into said extreme ultra violet light generating chamber when extreme ultra violet light is generated;a driver laser for emitting a laser beam;a window provided to said extreme ultra violet light generating chamber, for passing the laser beam into said extreme ultra violet light generating chamber;laser beam focusing optics including an optical element provided in said extreme ultra violet light generating chamber, said laser beam focusing optics focusing the laser beam emitted from said driver laser onto a trajectory of the target material injected into said extreme ultra violet light generating chamber to generate plasma;extreme ultra violet light collecting optics for collecting and outputting extreme ultra violet light radiated from said plasma;a cooling channel for supplying cooling water to said window and/or said optical element;a temperature sensor provided at a cooling water back-flow position of said cooling channel, for detecting a temperature of back-flow cooling water; anda processing unit for obtaining an amount of waste heat carried by the cooling water based on the temperature detected by the temperature sensor when extreme ultra violet light is generated, and determining deterioration of said window and/or said optical element based on the amount of waste heat. 21. The extreme ultra violet light source apparatus according to claim 20, wherein said driver laser includes a pre-pulse laser and a main pulse laser, and said window and said laser beam focusing optics are provided for each of said pre-pulse laser and said main pulse laser. 22. The extreme ultra violet light source apparatus according to claim 20, wherein said processing unit outputs a control signal for stopping emission of the laser beam to said driver laser when determining that said window and/or said optical element is deteriorated. 23. The extreme ultra violet light source apparatus according to claim 20, further comprising:a first energy detector for detecting energy of the laser beam, which has been dispersed without being applied to the target material after being focused by said laser beam focusing optics, when extreme ultra violet light is not generated,wherein said processing unit determines deterioration of said window and/or said optical element based on the energy of the laser beam detected by said first energy detector when extreme ultra violet light is not generated. 24. The extreme ultra violet light source apparatus according to claim 23, further comprising:a second energy detector provided outside said extreme ultra violet light generating chamber, for detecting energy of the laser beam before passing through said window,wherein said processing unit determines deterioration of said window and/or said optical element based on the energy of the laser beam detected by said first energy detector and the energy of the laser beam detected by said second energy detector when extreme ultra violet light is not generated. 25. The extreme ultra violet light source apparatus according to claim 23, wherein said first energy detector is provided outside said extreme ultra violet light generating chamber, and a concave mirror for guiding the laser beam, which has been dispersed without being applied to the target material, to said first energy detector is provided inside said extreme ultra violet light generating chamber. 26. The extreme ultra violet light source apparatus according to claim 25, further comprising:a protection housing provided in the extreme ultra violet light generating chamber,wherein said concave mirror is stored capable of being taken out in said protection housing, and is inserted into an optical path of the laser beam at measurement of said first energy detector. 27. The extreme ultra violet light source apparatus according to claim 23, wherein the target material supply unit does not inject the target material while said driver laser emits the laser beam, when said extreme ultra violet light is not generated. 28. The extreme ultra violet light source apparatus according to claim 23, wherein timing of injecting the target material controlled by the target material supply unit and timing of emitting the laser beam controlled by said driver laser are adjusted so as not to apply the laser beam to the target material, when said extreme ultra violet light is not generated. 29. The extreme ultra violet light source apparatus according to claim 23, wherein the trajectory of the target material controlled by the target material supply unit and an optical path of the laser beam controlled by said driver laser are adjusted so as not to apply the laser beam to the target material, when said extreme ultra violet light is not generated. 30. The extreme ultra violet light source apparatus according to claim 20, further comprising:a shielding means for shielding materials and electromagnetic waves generated in said extreme ultra violet light generating chamber to protect said window and said optical element provided in said extreme ultra violet light generating chamber when extreme ultra violet light is generated.
	claims	1. Screen for converting X-rays into light photons, comprising: a panel and an active layer for the conversion of X-rays into light photons located on one face of the panel, wherein the panel comprises a rigid foam plate that is transparent to X-rays, a first layer of composite material located on a first face of the rigid foam plate and a second layer of composite material located on a second face of the rigid foam plate, parallel to said first face. 2. Screen according to claim 1 , further comprising a framework located on the circumference of the rigid foam plate. claim 1 3. Screen according to claim 2 , wherein said framework comprises a matrix or glass or carbon fibres set in resin. claim 2 4. Screen according to claim 1 , wherein the rigid foam is a high density foam and the composite material comprises a matrix of glass or carbon fibres set in resin. claim 1 5. Screen according to claim 1 , further comprising: claim 1 a frame located around the face of the panel on which the active layer is located, so that the active layer is located within the interior of the frame. 6. Screen according to claim 5 , further comprising a layer of tungsten located between the panel and the active layer. claim 5 7. Radiological device comprising a screen for converting X-rays into light photons, including a conversion screen according to any of claims 1 - 6 .
059998946	abstract	In a method for the analysis of process data of an industrial plant, in particular a power station plant, parts of the plant are automatically controlled for the purpose of information compression, filtering and diagnosing faults in good time. Features characterizing the plant process are first prescribed and parameters relevant to the plant process are provided. Subsequently, the presence of each feature is checked for each plant part by using the parameters. Correlations between combinations of plant parts or features are determined by using features common to various plant parts and plant parts common to various features. Next, the plant parts and/or the features are represented as information elements positioned in such a way that the distance between two information elements in each case represents the degree of their correlation.
043022901	description	DESCRIPTION OF A PREFERRED EMBODIMENT Referring now to the drawing, and, more particularly, to FIG. 1 thereof, there is depicted therein a head equipment support structure system, hereinafter referred to by the acronym HESS, that is generally designated therein by reference numeral 10, and which embodies a construction in accordance with the present invention. In accord with the illustration of FIG. 1, the intent is to show therein the HESS 10 in cooperative association with a reactor vessel, and with the latter being generally designated in FIG. 1 through the use of the reference numeral 12. Within the latter reactor vessel 12 is housed the nuclear reactor (not shown) of a nuclear power generation system. Inasmuch as the manner in which the nuclear reactor (not shown) is housed within the reactor vessel 12 is well-known to those skilled in this art, it is not deemed necessary for purposes of acquiring an understanding of the present invention to set forth a description thereof herein, or to include illustrations thereof in the drawing. Likewise, the manner in which the reactor vessel 12 is operatively connected to the other components (not shown) which taken together collectively comprise a conventionally constructed nuclear power generation system is well-known to those skilled in the art. Accordingly, it is not deemed necessary to include a description thereof herein, or an illustration thereof in the drawing, particularly in view of the fact that it is only indirectly related to the subject matter of the present invention. However, should a description or illustration thereof be sought, reference may be had for this purpose to the teachings thereof that are readily available in the prior art. With further reference to FIG. 1, as shown therein, the HESS 10, in a manner which will be described more fully hereinafter, is suitably supported relative to the reactor vessel 12 so as to be located directly above the latter and in spaced relation thereto, while yet still being positioned so as to be operatively connectable thereto. To this end, the HESS 10 is located so as to be capable of performing the following functions; namely, that of a cable support structure, that of a missile shield, and that of a ductwork support structure. More specifically, the HESS 10 is intended to be operative to effect the separation and support of the electrical cabling associated with the operation of the nuclear reactor (not shown) that is housed within the reactor vessel 12. As regards the matter of a missile shield, preselected segments of the internal structure of the HESS 10 are designed to be capable of absorbing the kinetic energy of any missiles that strike thereagainst, such as to be operative in the manner of a missile shield. Lastly, the ductwork that is needed for purposes of achieving the requisite cooling of certain equipment associated with the operation of the nuclear reactor (not shown) housed in the reactor vessel 12 is designed to be mountable in supported relation on the HESS 10. Proceeding now with a description of the nature of the construction and the mode of operation of the HESS 10, reference will be had for this purpose, particularly to FIGS. 2 and 3 of the drawing. In accord with the preferred embodiment of the invention, the HESS 10 includes a multiplicity of decks 14. The latter decks 14 are suitably disposed so as to bear a substantially horizontal orientation when the HESS 10 is positioned in supported relation above the reactor vessel 12. The number of decks 14 with which the HESS 10 is provided may vary. However, preferably in order to effect the separation and support of the electrical cabling emanating from the reactor vessel 12, the HESS 10 embodies up to four decks 14. Since all of the decks 14 are substantially similar, if not virtually identical, in construction, it is not deemed necessary for purposes of obtaining an understanding of the present invention to describe each of the decks 14 in detail herein. Rather, the discussion will be limited in this regard to a description of the lower and upper decks 14' shown in FIGS. 2 and 3, respectively, that are usable in a HESS, and which represent the best mode embodiment thereof. With reference first to the matter of the upper deck 14', as shown in FIG. 3, the latter includes a pair of structural steel frame members 16, that extend in spaced relation to each other so as to in effect define two of the sides of the upper deck 14'. Each of the members 16 at the opposite ends thereof is provided with an overhang portion 18. The latter portions 18 are suitably configured so as to be capable of being cooperatively associated with appropriate portions of the walls 20 of the concrete structure within which, in accordance with customary practice, the reactor vessel 12 is suitably emplaced. In a manner well-known to those skilled in this art, the concrete walls 20 are located in surrounding relation to the reactor vessel 12 such that the steam generators (not shown) to which the reactor vessel 12 is operatively connected, are positioned on the other side of the walls 20. Further, the members 16 are constructed so as to be of sufficient length that they are capable of spanning the space which exists between opposing ones of the walls 20. Moreover, as best understood with reference to FIG. 3, the members 16 preferably are provided with a suitable number of reinforcing portions 22, which function to insure that the members 16 possess the strength required thereby in order to enable them to serve as support members for an entire HESS when the latter is emplaced above the reactor vessel 12. Continuing with a description of the upper deck 14' depicted in FIG. 3, the latter further includes a floor-like surface 24 which extends between the members 16 and is suitably interconnected thereto adjacent to the bottoms thereof, as viewed with reference to FIG. 3 through the use of any suitable conventional method (not shown) of effecting the interconnection of members. The floor-like surface 24 is preferably constructed of steel plates. Further, the surface 24 has provided therein suitable openings (not shown) through which the electrical cabling that emanates from the reactor vessel 12 is routed, as will be more fully described hereinafter. A plurality of termination panels 26 are suitably supported either on the floor-like surface 24 or attached to the walls 20 such that the panels 26 define planes that extend perpendicular to the plane of the surface 24. More specifically, in accord with the illustrated embodiment of the invention, the surface 24 has at least two such termination panels 26 mounted thereon. To this end, there is one such panel 26 mounted adjacent to the edge of one of the sides of the surface 24 that extends at right angles to the members 16, while the other such panel 26 is supported in adjacent relation to the other side of the surface 24, which extends perpendicularly to the members 16. Each of the termination panels 26 has a multiplicity of openings 28 formed therein, each of which is suitably configured and dimensioned so as to be capable of receiving one end of an electrical cable 30 therewithin in supported relation thereto such that the aforesaid end of the cable 30 is accessible for interconnection with the end of another cable that is mounted in supported relation relative to a wall 20 in a manner that is yet to be described. With further regard to the structure illustrated in FIG. 3, the surface 24 functions additionally as a support for a multiplicity of cables 30 that are routed through the openings (not shown) with which the surface 24 is suitably provided as has been described previously hereinabove. Once the cables 30 pass through the openings (not shown) they are suitably separated, i.e., segregated, and are fed to a corresponding one of the openings 28 that is provided for this purpose in a corresponding one of the termination panels 26. Thus, it can be seen from the above that the decks 14' are capable of being employed for purposes of assisting in the effectuation of the separation of the electrical cabling that emanates from the reactor vessel 12 and for purposes of providing support for the lengths of cables 30 as the latter extend across the length of the surface 24 and eventually terminate at the panels 26. The cables 30 can be alternately supported by a multiplicity of cableways 32 as shown in FIG. 3 spanning the reactor vessel and supported from the surface 24. These cableways 32 are channel-like and are mounted so as to extend on a diagonal between the members 16. The cableways 32 are suitably spaced one from another such that electrical cables 30 can be routed therebetween and into the cableways 32. The cableways 32 are structurally adequate to support the cables. Once the cables enter into the cableways 32 they are suitably separated, i.e., segregated and fed to a corresponding one of the openings 28 that is provided for this purpose in a corresponding one of the termination panels 26. Thus, it can be seen from the above that the cableways 32 are capable of being employed for purposes of assisting in the effectuation of the separation of the electrical cabling that emanates from the reactor vessel 12 and for purposes of providing support for the cables 30 as the latter extend across to the surface 24 and eventually terminate at the panels 26. In accord with the illustrated embodiment thereof shown in FIG. 3 the cableways 32 embody a configuration which resembles substantially two hollow boxes arrayed vertically. However, it is to be understood that the means of supporting the cables horizontally, i.e., on decks 14' or cableways 32 could embody some other configuration or form of construction without departing from the essence of the invention as long as the functions desired to be performed thereby, i.e., that of cable separation and support, are maintained. Lastly, suitable walk areas are provided by the surface 24 such that access can be had by personnel for purposes of effecting repairs and/or required maintenance on the equipment that is supported on the surface 24 or in sufficiently close proximity thereto so as to be reachable therefrom. Completing the description of the upper portion of the HESS shown in FIG. 3, the latter is preferably also provided with additional sections of floor-like surfaces, i.e., those designated therein by the reference numerals 34 and 36, respectively. The surfaces 34 and 36 are each suitably dimensioned so as to extend between the members 16. Moreover, the surfaces 34 and 36 are each suitably interconnected to the members 16 through the use of any suitable conventional form of means (not shown) operable for effecting the interconnection of adjoining surfaces. The surfaces 34 and 36 are preferably constructed of steel. The surfaces 34 and 36 are intended to perform a dual function. Namely, they serve as a support surface for additional pieces of equipment, to which further reference will be had hereinafter that are required in the operation of the nuclear reactor (not shown) housed within the reactor vessel 12. To this end, the surfaces 34 and 36 are each configured so as to be capable of receiving the desired pieces of equipment thereon in supported relation thereto. The other function performed by the surfaces 34 and 36 is that of providing a walkway for purposes of enabling access to be had to pieces of equipment by personnel desirous of conducting repair and/or maintenance operations thereon. As concerns the latter function, each of the surfaces 34 and 36 is in accord with the preferred embodiment of the invention equipped with suitably placed ladders (not shown) through which access may be gained to the surfaces 34 and 36 by the aforementioned personnel. Furthermore, for purposes of insuring the safety of the personnel using the surfaces 34 and 36 as walkways, suitable guardrails (not shown) are preferably provided around the peripheries of these surfaces 34 and 36. Turning now to a consideration of the lower deck 14' that is shown in FIG. 2 of the drawing, the latter embodies a form of construction that is similar to that of the upper deck 14' that has been described at length hereinabove. Accordingly, it is not deemed necessary for purposes of obtaining an understanding of the nature of the construction of the lower deck 14' of FIG. 2 to also describe the latter at length herein. Rather, it is deemed adequate to merely set forth below a brief descriptive summary of the nature of the construction thereof. To this end, the lower deck 14' includes a floor-like surface 38 suitably configured and dimensioned so as to be receivable within the area defined by the walls 20. In accord with the preferred embodiment of the invention, the surface 38 is constructed of steel and in appearance resembles a grate. Suitably supported in the central portion of the surface 38 are a multiplicity of channel-like cableways, i.e., dual compartment trays, 40. The latter cableways 40 are identical in construction and in function to the cableways 32 that were described above previously in connection with the discussion of the upper deck 14'. Namely, the cableways 40, like the cableways 32, function in the manner of a separation and support means for some of the electrical cables 30. That is, the latter cables 30 are suitably routed through openings (not shown) that have been provided for this purpose in the surface 38, and thereafter are received within the appropriate cableways 40 whereupon they are made to extend the entire length thereof. In addition, the ends of the cables 30 are received in suitable openings (not shown) provided for this purpose in the termination panels 42. The latter panels 42, the number which is determined according to the need therefor, are suitably positioned in mounted relation on the surface 38 adjacent to the edge of preselected ones of the sides thereof. In accord with the best mode embodiment of the invention, there would preferably be provided four such panels 42 arranged in two sets of two each. Like the panels 26 mounted on the surface 24 of upper deck 14', the panels 42 may be mounted on the surface 38 through the use of any suitable conventional form of mounting means (not shown). Lastly, note is taken of the fact that the panels 42 and the openings (not shown) formed therein are identical to the panels 26 and the openings 28 therein, respectively, both as regards the construction thereof and the mode of operation thereof. Like the surface 24 of the upper deck 14', the surface 38 of lower deck 14' is designed to function both in the manner of a support for equipment that it is desired to have located thereon and in the manner of a walkway whereby personnel may gain access to the equipment supported by the surface 38 or that located in sufficiently close proximity thereto so as to be accessible therefrom. To this end, surface 38 is equipped with strategically placed access ladders (not shown) whereby access may be gained to the surface 38. In addition, in accord with the preferred embodiment of the invention, the surface 38 is also equipped with suitably located guardrails (not shown) which are designed to afford protection to personnel utilizing the surface 38 as a walkway. The aforementioned ladders are guardrails with which the surface 38 is equipped have been omitted in the drawing in the interest of maintaining clarity of illustration therein, and because they are of conventional construction and are being employed in a manner well known to all. With further reference to FIG. 2 of the drawing, there is illustrated therein the upper end of a plurality of elements, each designated by the reference numeral 44, from which emanate a plurality of individual cables 30. The elements 44 appearing in FIG. 2 are intended to depict what is known to those skilled in this art as control element drive mechanisms. Briefly stated, the latter are devices that communicate with the interior of the reactor vessel 12, and, more specifically, with the nuclear reactor (not shown) housed therein. Moreover, these devices 44 are utilized in the exercise of the controlling of the operation of the nuclear reactor (not shown). For a fuller discussion of the nature of the construction and the mode of operation of the control element drive mechanisms 44 reference may be had to the teachings of the prior art. However, for purposes of the present invention, it is sufficient to merely note that the latter mechanisms 44 function as a source of the electrical cabling, i.e., the individual electrical cables 30 to which much reference has been had hereinbefore. Referring next to FIG. 4 of the drawing, there is illustrated therein on an enlarged scale the end portion of one of the control element drive mechanisms 44 that appears in FIG. 2 and to which reference has been had above. More specifically, there is depicted in FIG. 4 the upper end of a shroud 46 within which the operating components (not shown) of the control element drive mechanism 44 are encased. In accord with the preferred embodiment thereof, the shroud 46 embodies a substantially cylindrical configuration. Projecting outwardly from the end cap 48 of the shroud 46 and substantially at the center thereof is a vent 50. In addition, as shown in FIG. 4, a pair of rigid, pipe-like members 52, 54 also project outwardly of the end cap 48. In accord with the illustrated embodiment of the shroud 46, the members 52, 54 are located outwardly of and on opposite sides of the vent 50. Obviously, however, other forms of arrangements of these components could be utilized without departing from the essence of the invention. Likewise, a greater number of pipe-like members 52, 54 could be provided, if deemed desirable, without departing from the essence of the present invention. As best understood with reference to FIG. 4 of the drawing, each of the pipe-like members 52, 54 is designed to have one end of an electrical cable 30 connected thereto. To this end, in accord with the best mode embodiment of the invention, each electrical cable 30 is preferably encased in a flexible conduit. For ease of illustration in the drawing, the reference numeral 30 has been applied to the flexible conduit within which there is located one of the electrical cables 30 that emanates from the control element drive mechanism 44. As regards more specifically the aforedescribed flexible conduit, the latter, in actuality, in accord with the best mode embodiment of the invention, consists of a flexible conduit with outer braid. For a purpose that will be discussed more fully hereinafter, it is mandatory that the flexible conduit 30 with the electrical cable located therein be capable of flexing a specified amount, i.e., be capable of moving a specified distance in a horizontal direction from a designated vertical plane. The subject flexible conduit 30 with outer braid possesses the capability of fulfilling such a requirement. Continuing with a description of the structure depicted in FIG. 4, the end of the flexible conduit 30 which has an electrical cable encased therewithin is connected to a corresponding end of one of the pipe-like members 52, 54 through the use of a conventional form of connector 56, 58 respectively. Any suitable form of known connector means such as a threaded-type connector or a snap-lock type connector, etc., may be employed for purposes of effecting the interconnection of an end of a flexible conduit encased electrical cable 30 to the end of the pipe-like members 52, 54. It should be noted that what is important here is the concept of effectuating a rigid interconnection between the flexible conduit encased electrical cable 30 and the pipe-like members 52, 54 and not the specific connector means that is employed for this purpose. Rather, as noted above, any suitable connector means may be utilized for this purpose. Lastly, FIG. 4 contains an illustration of one of the flexible conduit encased electrical cables 30 being clamped by means of a conventional form of clamp 60 to a support member 62 through the use of a fastener 64. The point of attachment of the clamp 60 to the flexible conduit 30 is preselected so as to insure that the latter is capable of flexing in a horizontal direction to the required degree. Namely, as will be described herein more fully subsequently, the flexible conduit 30 must be permitted to undergo a sufficient amount of flexing in the horizontal direction relative to the vertical plane defined by, and in accord with the illustration of FIG. 4, the pipe-like member 54. Thus, for this purpose, the clamp 60 must be located in suitably spaced relation to the point of connection of the end of the flexible conduit 30 to the pipe-like member 54 to enable the aforedescribed flexing to take place. With reference again to FIG. 1 of the drawing, there is depicted therein the ductwork to which mention was made previously hereinabove. More specifically, the upper portion of the HESS 10 is intended to function as a support for an exhaust plenum 66. Moreover, in accord with the illustration of FIG. 1, the latter plenum 66 has suitably connected thereto an exhaust elbow 68, the other end of which is connectable to a suitable exhaust outlet (not shown) in a manner and for a purpose that is well-known to those skilled in this art and, accordingly, need not be described herein. In addition, a plurality of exhaust conduits 70, 72 and 74 each have one end thereof connected to the plenum 66 while the other end thereof is cooperatively associated with the reactor vessel 12 such as to comprise an exhaust circuit from the latter through the plenum 66 to the afore-referenced exhaust outlet (not shown). The aforedescribed ductwork consisting of the plenum 66, the exhaust elbow 68, and the exhaust conduit 70, 72 and 74 are suitably mounted on the HESS 10 so as to form an integral part thereof whereby when the HESS 10 is removed from its illustrated position of FIG. 1 relative to the reactor vessel 12, the aforesaid ductwork is removable therewith. Accordingly, the ends of the exhaust elbow 68 and the exhaust conduit 70, 72 and 74 that are not connected to the plenum 66 are provided with suitable disconnecting means (not shown) to enable the accomplishment of the aforementioned removal of the HESS 10 from the position occupied thereby in FIG. 1. Completing the description of the HESS 10, as best understood with reference to FIG. 1 of the drawing, the HESS 10, in accord with the best mode embodiment of the invention, is provided with a vertically extending leg-like column 76 located at each of the four corners thereof, only three such columns 76 being visible in FIG. 1. The columns 76 are rigidly affixed to one or more of the decks 14 so as to extend substantially perpendicular thereto, through the use of any suitable conventional form of fastening means (not shown). The principal function of the columns 76 is to serve as support legs for the HESS 10 when the latter occupies its laydown position; namely, when the HESS 10 is removed from emplacement above the reactor vessel 12, i.e., removed from the position thereof depicted in FIG. 1. A description will now be had of the mode of operation of the HESS 10. Normally, the latter will be found emplaced above the reactor vessel 12 in the manner depicted in FIG. 1. When so positioned, the HESS 10 is suitably supported by virtue of the engagement of the overhang portions 18 of the frame member 16 with appropriate portions of the walls 20. As such, the multiple decks 14 of the HESS 10 lie within the area defined within the walls 20. However, if it is deemed desirable, provision could be made to have portions of others of the decks 14 engage support members provided for this purpose at intermediate locations along the length of the walls 20. Two such portions identified through the use of the reference numeral 78 have been shown in FIG. 2 by way of exemplification. With further regard to the mode of operation of the HESS 10, the electrical cables 30 emanating from the control element drive mechanisms 44 are suitably routed through the openings (not shown) that are formed for this purpose in the various floor-like surfaces, e.g., 24, 38, etc., with which each of the multiple decks 14' are provided. To this end, the electrical cables 30 upon leaving the mechanisms 44 are suitably separated such as to be routed to the appropriate one of the multiple decks 14'. In this regard, the separation of the electrical cables is accomplished so as to insure that the desired isolation between primary cables and the cables that function as backup thereto is achieved. By way of illustration in this regard, there are various types of cables that extend from the mechanisms 44. Namely, there is power cabling, reed switch position transmitter cabling and in-core instrumentation cabling. Moreover, the reed switch position transmitter cabling encompasses two sets of such cabling between which it is desired to effect isolation. In accord with the best mode embodiment of the invention, each of these various types of cables is routed in a preselected manner to effect the desired separation and/or isolation therebetween. In this connection, not only are the cables 30 sorted for routing to individual ones of the decks 14', but they also are sorted so as to be receivable within a given one of the cableways 32 or 40 such that they can be fed properly to the appropriate opening in the proper termination panels 26 or 42. To summarize, the separation of the electrical cables 30 is achieved by the routing thereof to the appropriate location, i.e., to the proper deck 14', the proper cableway 32 or 40 and the proper opening in the proper termination panel 26 or 42. Continuing with the description of the mode of operation of the HESS 10, once the electrical cables 30 have been routed therethrough in the manner described above, should there be a need to effect repairs therein necessitating the removal of one or more cables 30 therefrom, this task can be easily accomplished. Namely, with the HESS 10 emplaced above the reactor vessel 12 in the manner depicted in FIG. 1, the removal of a cable 30 from the HESS 10 can be accomplished simply by disconnecting both ends of the appropriate cable 30 and disconnecting the clamp 60 if such should be employed, while thereafter effecting the withdrawal of the latter cable 30 from the appropriate deck 14', and termination panel 26 or 42. Similarly, utilizing the floor-like surfaces of the decks 14' as walkways, personnel may effect the afore-described withdrawal of the cable 30 as well as accomplish repairs and/or normal maintenance on any of the pieces of equipment accessible therefrom. In accord with the best mode embodiment of the invention, the ends of the cables 30 that are supported in the termination panels such as the panels 26 and 42, are designed to be connected to the ends of other cables, such as those designated by reference numeral 80 in FIG. 2, that are supported by the walls 20. Moreover, the connection of the ends of the cables 30 with the ends of the cables 80 preferably is effected through the use of suitable means (not shown) similar to the connector means 56, 58 depicted in FIG. 4. Finally, for purposes of spanning the gaps that exist between the edges of the decks 14 and the walls 20, suitable tray-like members may be provided. Preferably, such tray-like members would extend between the edges of the decks 14' and the walls 20 and would provide a planar support for the electrical cabling extending therebetween. In order to facilitate the removal of the HESS 10 from the position occupied thereby in FIG. 1, such tray-like members preferably would either take the form of a tray that is pivotably affixed to the HESS 10 so as to be pivotable relative thereto between an extended and a retracted position, or such tray-like members would take the form of a tray which is retractably mounted on the walls 20 so as to be movable relative thereto between an extended and a retracted position. Referring again to the matter of the clamp 60 depicted in FIG. 4, the need to allow for movement by the electrical cables 30 stems from the fact that allowance must be made for the possibility that seismic occurrences may taken place that will affect the HESS 10; namely, that will cause the latter, and, more particularly, the components encompassed thereby to move. To this end, this is the reason why flexible conduit has been proposed for use in the HESS 10. Other means such as telescoping conduits, etc., had been proposed for use, but generally have been found to be unacceptable from the standpoint of their ability to fulfill the requirement that allowance be made for movement of the cables 30 occasioned by the occurrences of seismic events. The fact that the HESS 10 additionally is operative as a support for auxiliary-type equipment such as the ductwork encompassing the plenum 66, elbow 68 and conduit 70, 72 and 74 shown in FIG. 1 has been discussed above in detail, and accordingly, it is not deemed necessary to make further reference thereto at this point. However, mention was made above to the fact that the HESS 10 is operative in the manner of a missile shield. To this end, the multiple decks 14 are operative to absorb the kinetic energy possessed by missiles striking thereagainst as the latter are propelled away from the reactor vessel 12. Moreover, suitable steel plates may be mounted in supported relation to the upper deck 14 as a further barrier to such missiles. These steel plates would be designed to form an integral part of the HESS 10 so as to be movable as unit therewith. In summary, the decks 14 and the aforesaid steel plates are capable of providing a missile shield that is functionally as effective as the concrete missile shield that has commonly been employed heretofore. However, the former possesses a major advantage over the latter, both in terms of the ease of removal thereof as well as in the time and cost involved in effecting the placement thereof. A discussion has been had previously hereinabove of the importance of such factors. Finally, in the event that it becomes necessary to remove the HESS 10 from the position thereof shown in FIG. 1 relative to the reactor vessel 12, i.e., move the HESS 10 to a laydown position, this is easily accomplished. Namely, in accord with the best mode embodiment of the invention, the HESS 10 is preferably provided with a plurality of strategically located lifting lugs (not shown). Through the engagement of the latter lugs (not shown) by suitable lifting means, the entire HESS 10 may be raised from engagement with the walls 20, and moved to another position, i.e., a laydown position, whereby unobstructed access may be had to the reactor vessel 12. Obviously, before the HESS 10 would be raised, as described above, the cables 30 and the conduits 70, 72 and 74 as well as any other elements that may require such action would be suitably disconnected. When moved to its laydown position, the HESS 10 is made to rest on the four columns 76. Thus, in accordance with the present invention, there has been provided a new and improved form of support structure system that is particularly suited for use in cooperative relation with a reactor vessel in a nuclear power generation system for purposes of providing support for the head equipment associated therewith. Moreover, the subject head equipment support structure system is operable in the manner of a cable support structure. In addition, in accord with the present invention, a head equipment support structure system is provided that in addition is operable in the manner of a missile shield. Further, the head equipment support structure system is also operable in the manner of a ductwork support structure. Additionally, in accordance with the present invention, a head equipment support structure system is provided that is characterized in the ease with which it may be repositioned for purposes of effecting the removal of the head of the reactor vessel. Also, the head equipment support structure system of the present invention is characterized in the fact that it is equally applicable for use in new installations as well as retrofit applications. Furthermore, in accord with the present invention, a head equipment support structure system that renders it possible through the use thereof to attain measurable cost-savings and time-savings as compared to the costs and times associated with the use of prior art forms of support structure systems. While one one embodiment of our invention has been shown it will be appreciated that modifications thereof, some of which have been alluded to hereinabove, may still be readily made thereto by those skilled in the art. We, therefore, intend by the appended claims to cover the modifications alluded to herein as well as all other modifications, which fall within the true spirit and scope of our invention.
	claims	1. A method for determining a convex or a concave part of a sample by detecting charged particles emitted from the sample by scanning a charged particle beam, comprising:changing a focus position of the charged particle beam;forming a plurality of line profiles, each line profile including a peak which indicates an edge part located between a convex or concave part of the sample and the adjacent surface, at the different focus positions;calculating, for each of the plurality of line profiles, a first width on one side of the peak of the line profile and a second width on the other side of the peak of the line profile; anddetermining that the convex part is located on the one side of the edge part when a minimum value of the first widths corresponds to a shorter focus length than a focus length corresponding to a minimum value of the second width. 2. A method for determining a convex or a concave part of a sample by detecting the charged particles emitted from a sample by irradiation of a charged particle beam emitted from a charged particle source and determining the uneven surface of the sample based on the detection of the charged particles, wherein:a plurality of determinations are performed by the method for determining a convex or a concave part of a sample according to claim 1. 3. A sample measuring method for detecting the charged particles emitted from a sample by irradiation of a charged particle beam emitted from a charged particle source, wherein the sample is measured according to the focal position when a value subjected to measurement used for the determination of a convex or a concave part of a sample according to claim 1 becomes minimum. 4. A charged particle beam apparatus, comprising:a charged particle source,a scanning deflector for scanning a charged particle beam emitted from the charged particle source on a sample,a lens for changing a focus position of the charged particle beam emitted from the charged particle source,a detector for detecting charged particles emitted from the sample, anda computing section for processing a signal based on the detected charged particles, wherein the computing section:forms a plurality of line profiles, each line profile including a peak which indicates an edge part located between a convex or concave part of the sample and the adjacent surface, at different focus positions;calculates, for each of the plurality of line profiles, a first width on one side of the peak of the line profile and a second width on the other side of the peak of the line profile; anddetermines that the convex part is located on the one side of the edge part when a minimum value of the first widths corresponds to a shorter focus length than a focus length corresponding to a minimum value of the second widths. 5. A computer readable storage medium comprising instructions for determining a convex or a concave part of a sample by detecting charged particles emitted from the sample by scanning of a charged particle beam, which when executed by a computer cause the computer to perform a method comprising steps of:changing a focus position of the charged particle beam;forming a plurality of line profiles, each line profile including a peak which indicates an edge part located between a convex or concave part of the sample and the adjacent surface, at the different focus positions;calculating, for each of the plurality of line profiles, a first width on one side of the peak of the line profile and a second width on the other side of the peak of the line profile; anddetermining that the convex part is located on the one side of the edge part when a minimum value of the first widths corresponds to a shorter focus length than a focus length corresponding to a minimum value of the second width. 6. The charged particle beam apparatus according to claim 4, wherein:a template for specifying the convex or concave part of the sample and the measured results of the convex or concave part are compared to determine a state of the convex or concave part of the sample. 7. The charged particle beam apparatus according to claim 4, further comprising a means for displaying by superimposing the convex or concave part information of the sample on the photographed image as a profile waveform. 8. The charged particle beam apparatus according to claim 4, wherein the obtained convex or concave part information is used to specify the position of the sample. 9. The charged particle beam apparatus according to claim 4, wherein an image or value obtained when autofocusing is used to calculate a signal or profile waveform used for determination. 10. The charged particle beam apparatus according to claim 4, wherein a convex or concave part of the scanned portion is determined, and pattern matching is performed according to the results of the convex or concave part determination. 11. The method according to claim 1, wherein the detection of the charged particles is performed on at least two focal positions, and a signal based on the detected charged particles is used to determine the convex or the concave part of the sample. 12. A method for determining a convex or a concave part of a sample by detecting charged particles emitted from the sample by scanning a charged particle beam, comprising:changing a focus position of the charged particle beam;forming a plurality of line profiles, each line profile including a peak which indicates an edge part located between a convex or concave part of the sample and the adjacent surface, at the different focus positions;calculating, for each of the plurality of line profiles, a first width on one side of the peak of the line profile and a second width on the other side of the peak of the line profile; anddetermining that the concave part is located on the one side of the edge part when a minimum value of the first widths corresponds to a longer focus length than a focus length corresponding to a minimum value of the second width. 13. The method according to claim 12, wherein the detection of the charged particles is performed on at least two focal positions, and a signal based on the detected charged particles is used to determine the convex or the concave part of the sample. 14. A method for determining a convex or a concave part of a sample by detecting the charged particles emitted from a sample by irradiation of a charged particle beam emitted from a charged particle source and determining the uneven surface of the sample based on the detection of the charged particles, wherein:a plurality of determinations are performed by the method for determining a convex or a concave part of a sample according to claim 12. 15. A sample measuring method for detecting the charged particles emitted from a sample by irradiation of a charged particle beam emitted from a charged particle source, wherein the sample is measured according to the focal position when a value subjected to measurement used for the determination of a convex or a concave part of a sample according to claim 12 becomes minimum. 16. A charged particle beam apparatus, comprising:a charged particle source,a scanning deflector for scanning a charged particle beam emitted from the charged particle source on a sample,a lens for changing a focus position of the charged particle beam emitted from the charged particle source,a detector for detecting charged particles emitted from the sample, anda computing section for processing a signal based on the detected charged particles, wherein the computing section:forms a plurality of line profiles, each line profile including a peak which indicates an edge part located between a convex or concave part of the sample and the adjacent surface, at different focus positions;calculates, for each of the plurality of line profiles, a first width on one side of the peak of the line profile and a second width on the other side of the peak of the line profile; anddetermines that the concave part is located on the one side of the edge part when a minimum value of the first widths corresponds to a longer focus length than a focus length corresponding to a minimum value of the second widths. 17. A computer readable storage medium comprising instructions for determining a convex or a concave part of a sample by detecting charged particles emitted from the sample by scanning a charged particle beam, which when executed by a computer cause the computer to perform a method comprising steps of:changing a focus position of the charged particle beam;forming a plurality of line profiles, each line profile including a peak which indicates an edge part located between a convex or concave part of the sample and the adjacent surface, at the different focus positions;calculating, for each of the plurality of line profiles, a first width on one side of the peak of the line profile and a second width on the other side of the peak of the line profile; anddetermining that the concave part is located on the one side of the edge part when a minimum value of the first widths corresponds to a longer focus length than a focus length corresponding to a minimum value of the second width. 18. The computer readable storage medium according to claim 17, wherein:a template for specifying the convex or concave part of the sample and the measured results of the convex or concave part are compared to determine a state of the convex or concave part of the sample. 19. The computer readable storage medium according to claim 17, the method further comprising displaying by superimposing the convex or concave part information of the sample on the photographed image as a profile waveform. 20. The computer readable storage medium according to claim 17, wherein the obtained convex or concave part information is used to specify the position of the sample. 21. The computer readable storage medium according to claim 17, wherein an image or value obtained when autofocusing is used to calculate a signal or profile waveform used for determination. 22. The computer readable storage medium according to claim 17, wherein a convex or concave part of the scanned portion is determined, and pattern matching is performed according to the results of the convex or concave part determination.
056065868	claims	1. An exposure method for transferring a pattern onto a substrate using synchrotron radiation, said method comprising: directly measuring a luminous intensity distribution of the synchrotron radiation, by a radiation detector, at least in a predetermined area where the pattern is to be transferred; determining a relationship between the luminous intensity distribution of the synchrotron radiation and an exposure amount distribution absorbed by the substrate, at least in the predetermined area where the pattern is to be transferred; and effecting an exposure operation while controlling a dose amount for respective positions in the predetermined area using the determined relationship. determining relationships between at least two accumulation currents of the source of the synchrotron radiation and an exposure amount distribution absorbed by the substrate, at least in a predetermined area where the pattern is to be transferred; and controlling a dose amount of the synchrotron radiation for respective positions in the predetermined area on the basis of the relationship and an accumulation current detected during an exposure operation. radiation detecting means for directly measuring a luminous intensity distribution of the synchrotron radiation, at least in a predetermined area where the pattern is to be transferred; means for determining a relationship between the luminous intensity distribution of the synchrotron radiation and an exposure amount distribution absorbed by the substrate, at least in the predetermined area where the pattern is to be transferred; and means for effecting an exposure operation while controlling a dose amount for respective positions in the predetermined area using the determined relationship. means for determining relationships between at least two accumulation currents of the source of the synchrotron radiation and an exposure amount distribution absorbed by the substrate, at least in a predetermined area where the pattern is to be transferred; and means for controlling a dose amount of the synchrotron radiation for respective positions in the predetermined area on the basis of the relationship and an accumulation current detected during an exposure operation. 2. A method according to claim 1, wherein the dose amount is controlled by changing a driving profile of a movable shutter for controlling the exposure operation. 3. A method according to claim 1, wherein the synchrotron radiation intensity distribution is detected upon the exposure operation, and the dose amount is also controlled on the basis of the detection of the synchrotron radiation intensity distribution. 4. A method according to claim 1, wherein an accumulation current of the source of the synchrotron radiation is detected upon the exposure operation, and the dose amount is also controlled on the basis of the detected accumulation current. 5. A method according to claim 1, wherein the exposure amount distribution is determined on the basis of a remaining film ratio of a resist material on a substrate in a test exposure operation. 6. A method according to claim 1, wherein the relationship is in the form of a proportional coefficient between the radiation intensity and the exposure amount as a function of position information in the predetermined area. 7. A method according to claim 1, wherein the position information includes coordinate information in the predetermined area. 8. A method according to claim 6, wherein the position information includes information relating to deviation from a predetermined position in the predetermined area. 9. An exposure method for transferring a pattern onto a substrate using synchrotron radiation, said method comprising: 10. An exposure apparatus for transferring a pattern onto a substrate using synchrotron radiation, said apparatus comprising: 11. An exposure apparatus for transferring a pattern onto a substrate using synchrotron radiation, said apparatus comprising:
	description	This application is a division of U.S. patent application Ser. No. 15/689,724 filed on Aug. 29, 2017, now U.S. Pat. No. 10,726,961, which is a division of U.S. patent application Ser. No. 14/064,291 filed on Oct. 28, 2013, now U.S. Pat. No. 9,779,840, the disclosures of which are hereby incorporated by reference in their entirety. This invention was made with Government support under Contract No. DE-NE0000583 awarded by the Department of Energy. The Government has certain rights in this invention. The following relates to the nuclear power generation arts, nuclear reactor safety arts, nuclear reactor control arts, and related arts. During normal operation of a nuclear reactor, the nuclear chain reaction (and hence the thermal output of the reactor) is controlled to maintain temperature and pressure of the coolant water in the reactor pressure vessel in a normal range. This control is achieved using mechanisms such as a control rods system, control of steam generator feed water flow, adjusting a concentration of soluble boron neutron poison in the coolant water, or so forth. The steam generator acts as the heat sink for the nuclear reactor. If a pipe rupture interrupts feed water flow into the steam generator or steam flow out of the steam generator, a feed water flow is lost, or so forth, this heat sinking is compromised. In such a loss of heat sinking event, pressure and temperature of the coolant in the reactor pressure vessel rise, and this must be counteracted to maintain the nuclear reactor in a safe condition. To this end, the control rods are inserted (scrammed) to extinguish the nuclear chain reaction. However, the reactor core continues to output residual heat due to radioactive intermediate products of the nuclear chain reaction. Passive cooling systems reliant on a closed-loop evaporation/condensation cycle may be employed to remove this residual heat. If the passive cooling systems are unable to keep up with the residual heat output from the shut-down reactor core, then the temperature and pressure of the coolant inside the pressure vessel will continue to rise. In this circumstance, an emergency core cooling system (ECCS) is brought online to depressurize the pressure vessel by venting (into a suitable condenser, or into the surrounding radiological containment structure, or other suitable structure) while maintaining sufficient water flow into the reactor pressure vessel to avoid exposure of the nuclear reactor core. In one disclosed aspect, an apparatus comprises: a pressurized water reactor (PWR) including a nuclear reactor core comprising fissile material disposed in a reactor pressure vessel also containing primary coolant water; a pressurizer integral with or operatively connected with the reactor pressure vessel and configured to control pressure in the reactor pressure vessel; and a decay heat removal system including a pressurized passive condenser and a pump driven by a turbine in which steam from the pressurizer drives the turbine and exhausts into the pressurized passive condenser and the pump is connected to suction water from the pressurized passive condenser into the reactor pressure vessel. A pressurizer power operated relief valve may be configured to control discharge of steam bypassing the turbine into the pressurized passive condenser to control pressure in the pressurizer. A pressurizer block valve may be configured to activate the decay heat removal system by opening to admit steam from the pressurizer to the turbine. The pump and the turbine may be mounted on a common shaft so that the shaft provides direct mechanical coupling via which the turbine drives the pump. In another disclosed aspect, a method operates in conjunction with a pressurized water reactor (PWR) including a nuclear reactor core comprising fissile material disposed in a reactor pressure vessel also containing primary coolant water, a pressurizer integral with or operatively connected with the reactor pressure vessel and configured to control pressure in the reactor pressure vessel, and a refueling water storage tank (RWST). The method comprises responding to a loss of heat sinking of the PWR by operations including driving a turbine using steam piped from the pressurizer, and driving a pump using the turbine to suction water from the RWST into the reactor pressure vessel. The driving of the pump may comprise providing a common shaft mechanically connecting the turbine and the pump whereby the driven turbine rotates the common shaft to drive the pump. The method may further include the operations of discharging steam piped from the pressurizer into a pressurized passive condenser, and connecting the suction side of the pump to both the RWST and the pressurized passive condenser wherein the driving of the pump also suctions water from the pressurized passive condenser into the reactor pressure vessel. In another disclosed aspect, a decay heat removal system operates in conjunction with a pressurized water reactor (PWR) including a nuclear reactor core comprising fissile material disposed in a reactor pressure vessel also containing primary coolant water and a pressurizer integral with or operatively connected with the reactor pressure vessel and configured to control pressure in the reactor pressure vessel. The decay heat removal system comprises: a pressurized passive condenser; a turbine; a pump driven by the turbine and connected to suction water from at least one water source into the reactor pressure vessel; and steam piping configured to deliver steam from the pressurizer to the turbine to operate the pump and to discharge the delivered steam into the pressurized passive condenser. The pump and the turbine may be mounted on a common shaft so that the shaft provides direct mechanical coupling via which the turbine drives the pump. The at least one water source may include a refueling water storage tank (RWST) disposed with the PWR in a radiological containment structure and/or the pressurized passive condenser. The steam piping may include a pressurizer power operated relief valve configured to control discharge of a portion of the delivered steam bypassing the turbine into the pressurized passive condenser to control pressure in the pressurizer. Disclosed herein are embodiments of an alternative safety function (ASF) system that provides a safety response that is more aggressive than reliance upon passive evaporation/condensation systems but which avoids the loss of coolant entailed in invoking the emergency core cooling system (ECCS). The disclosed ASF system provides an alternative mechanism to control the reactor coolant system (RCS) temperature and inventory thereby avoiding (at least in some postulated events) actuation of the ECCS with its concomitant reactor pressure vessel depressurization in the event of a loss of heat sinking due to an event such as a steam generator tube rupture (SGTR), a main steam line break, a feed water line break, or a loss of feed water. The disclosed ASF system also provides an alternative mechanism for reactor shutdown in the event that the primary reactor shutdown mechanism is ineffective. With reference to FIG. 1, a nuclear reactor island including a nuclear reactor 8 is diagrammatically shown. FIG. 1 diagrammatically represents the nuclear island by way of symbolic icons representing components of the island. The nuclear reactor 8 includes a nuclear reactor core 10 comprising fissile material disposed in a reactor pressure vessel (RPV) 12 containing (primary) coolant, typically in the form of purified water (H2O), comprising the reactor coolant system (RCS). Optionally, the coolant may include additives such as a soluble boron poison to assist in reactivity control. The nuclear reactor core 10 is typically constructed as UO2 pellets enriched in fissile 235U and loaded into fuel rods assembled into fuel assemblies which are in turn arranged to form the reactor core. The fuel rods may have various compositional arrangements, e.g. axial zones of different enrichment levels, inclusion of neutron poison pellets or compositional components, or so forth. Thermal power output due to the nuclear chain reaction in the reactor core 10 is regulated by operation of a control rods system 14 comprising control rods of a neutron-absorbing material controllably inserted into/withdrawn from the nuclear reactor core 10 by suitable control rod drive mechanisms (CRDM's). The diagrammatically indicated control rods system 14 is an internal system whose CRDMs are located inside the reactor pressure vessel 12; alternatively, external CRDMs may be employed along with suitable mechanical feedthroughs and/or pressure boundary extensions to couple into the RPV 12. The illustrative nuclear reactor 8 is of the pressurized water reactor (PWR) variety and includes an integral pressurizer 16 for controlling reactor coolant pressure inside the reactor pressure vessel 12; alternatively, an external pressurizer connected with the pressure vessel by suitable piping is contemplated. The illustrative PWR 8 is an integral PWR in which an internal steam generator 18 is disposed inside the reactor pressure vessel 12; alternatively, an external steam generator may be connected with the RPV 12 by suitable piping. The diagrammatically indicated internal steam generator 18 is a once-through steam generator (OTSG), but other steam generator designs, such as a helical steam generator, are also contemplated. Moreover, two or more steam generator units may be provided for redundancy. In the steam generator 18, (secondary) coolant water flow is placed into thermal communication with the primary coolant water to heat sink the PWR 8. The illustrative nuclear island also includes components of an emergency core cooling system (ECCS), including a pressurized boron storage tank 20 containing borated water (that is, water containing a high concentration of soluble boron poison) for injection into the reactor pressure vessel 12 as part of the ECCS response, a refueling water storage tank (RWST) 22 which provides a water reservoir for routine reactor refueling operations as well as providing a water supply for long-term decay heat removal after depressurization of the reactor pressure vessel 12. During normal operation (configuration not shown in FIG. 1), secondary coolant is injected into the steam generator 18 via a feed water inlet of the RPV 12, and steam is output from a steam outlet of the RPV 12. The secondary coolant flowing inside the steam generator 18 is heated by heat transfer from the primary coolant inside the RPV 12 to convert the feed water to steam. Although the secondary coolant water is in thermal communication with the primary coolant water, it is in fluid isolation from the primary coolant water. For example, in a tube-and-shell steam generator design, primary coolant flows in a shell surrounding tubes carrying the secondary coolant (or vice versa). The steam output from the steam outlet of the RPV 12 is suitably used to perform useful work, for example driving the turbine (not shown) of an electrical generator in the case of a nuclear power plant. The nuclear chain reaction in the nuclear reactor core 10 generates the thermal output that heats the primary coolant water, and the nuclear chain reaction is controlled by operation of the control rods system 14 and/or other control mechanisms such adjusting the concentration of soluble boron in the primary coolant water, adjusting the feed water flow rate or temperature, or so forth. In a loss of heat sinking event, closed-loop circulation of secondary coolant through the steam generator 18 is interrupted. In this event, an alternative safety function (ASF) system is activated. The ASF system includes a pressurized passive condenser (PPCND) 24 with condenser coils 26 and a passive cooling water tank 28. (Other thermal sinks are contemplated in place of the passive cooling water tank 28, such as a PPCND employing an air-cooled radiator with battery-operated fans). The ASF system further includes a turbine-driven pump 30 that is driven by a turbine 32. In the illustrative embodiment, the pump 30 and the turbine 32 are mounted on a common shaft 34 so that the shaft 34 provides direct mechanical coupling via which the turbine 32 drives the pump 30. In other contemplated embodiments (not illustrated), the turbine includes an electrical generator (i.e., a turbine-generator on a common shaft) and electricity from the generator of the turbine-generator system drives the pump. However, the illustrative embodiment including the pump 30 and turbine 32 on a common shaft 34 advantageously reduces the number of parts (eliminating, e.g. the generator and electrical connections between the generator and pump) and consequently is expected to be more robust and less prone to malfunction. The ASF system also includes associated connecting piping and valves. The simplified diagrammatic representation of FIG. 1 shows: a valve 35 on the drain of the RWST 22, a pressurized passive condenser (PPCND) level control valve 36, recirculation flow block valves 37, a steam turbine control valve 39, a back-pressure control valve 40 that controls charging flow to the reactor coolant system (RCS), a pressurizer power operated relief valve (pressurizer PORV) 42 providing pressurizer pressure control, boron storage tank valves 41, 43 controlling release of borated water from the boron storage tank 20, a pressurizer steam supply block valve 44, and check valves 34, 45, 46. The ASF system provides passive recirculation in the event of a loss of heatsinking event. In contrast to an evaporation/condensation recirculation system, the ASF system employs the pump 30 to drive closed-loop circulation of water, with the motive force for the pump 30 being provided by pressure in the pressurizer 16 via the turbine 32 and common shaft 34. The pressurizer 16 is at elevated pressure during a loss of heatsinking event due to the residual decay heat generated by the nuclear reactor core heating the primary coolant. A portion of this pressure is vented as steam to the turbine 32 which drives the pump 30, with the turbine exhausting to the pressurized passive condenser (PPCND) 24 via a turbine exhaust line 48. With continuing reference to FIG. 1 and with further reference to FIG. 2, during a loss of heatsinking event the nuclear island assumes the configuration shown in FIG. 1. In an operation 100, a trigger event is detected that causes activation of the ASF system. The trigger event may, for example, be detection of a steam generator tube rupture (SGTR), detection of a main steam line break, detection of a feed water line break, or detection of a loss of feed water. Such detection is preferably automatic, e.g. based on steam pressure read by a pressure sensor dropping below a trigger threshold or feed water flow read by a flow meter dropping below a trigger threshold. Additionally or alternatively, the trigger event may be a manual activation of the ASF system. Upon activation, in an operation 102 the pressurizer block valve 44 opens and the steam turbine control valve 39 throttles open admitting steam to the turbine 32 which provides motive force to the pump 30. The turbine 32 exhausts to the pressurized passive condenser 24 via the turbine exhaust line 48. In an operation 104, pressurizer pressure and hence RCS temperature (RCS is at saturation) is controlled by the pressurizer PORV 42 with the discharge sparged in the pressurized passive condenser 24 via a discharge line 50 and spargers 52. Once the pressurized passive condenser 24 exceeds the static head of the RWST 22 it is able to provide a suction source to the pump 30 (assuming sufficient level in the pressurized passive condenser 24). In one contemplated embodiment, the pressurized passive condenser 24 operates at approximately 100 psia and 325° F. (160° C.), although other condenser operating temperature/pressure conditions are also contemplated. The elevated temperature in the pressurized passive condenser 24 reduces the thermal stress associated with the steam discharge into the pressurized passive condenser 24 via the discharge line 50, which in one contemplated typical PWR is at a steam temperature of about 600° F. (315° C.). The steam is sparged into the liquid zone of the pressurized passive condenser 24 via the spargers 52 to further reduce thermal stresses on the pressurized passive condenser 24. The ASF system is a passive safety system, and requires no external power other than the pressure in the pressurizer 16 (which is inherently present in a loss of heatsinking event) and power for valves/controller circuits which is suitably supplied by a battery, alternator, nitrogen bottles providing pneumatic valve motive force, or so forth. Thus, the disclosed ASF system supports long-term operation without outside support. The pump 30 provides sufficient pressure to provide flow to the RCS to support decay heat removal and postulated LOCA inventory losses. The injection flow is controlled by the back-pressure control valve 40 to maintain level in the pressurizer. The condenser level control valve 36 provides closed loop supply of working fluid to facilitate long-term cooling capability. In the illustrative embodiment, the suction side of the pump 30 is connected with both the pressurized passive condenser 24 and the RWST 22 so that the pump 30 can suction water into the RPV 12 from one or both sources 22, 24. The two sources 22, 24 provide synergistic benefits: the pressurized passive condenser 24 receives steam from the RPV 12 via the turbine exhaust line 48 and the discharge line 50, so as to form a closed-loop recirculation path that can in principle operate indefinitely. On the other hand, the RWST 22 provides a large reservoir of water located inside the radiological containment structure with the PWR 8 that can be tapped to maintain decay heat removal in the event of a leak associated with the pressurized passive condenser 24. While the illustrative embodiment leverages both synergistic benefits by connecting the suction side of the pump 30 to both the RWST 22 and the pressurized passive condenser 24, it is contemplated to alternatively connect the suction side of the pump 30 to only one of these water sources (i.e. to only the RWST 22 but not the pressurized passive condenser 24; or, to only the pressurized passive condenser 24 but not the RWST 22). Additionally, in the illustrative embodiment of FIG. 1, the boron storage tank 20 is injected into the RPV 12 via the pump 30. In this embodiment, the pump 30 can take suction from the boron storage tank 20 when the valves 41, 43 are both open (for alternative reactor shutdown) and provides discharge pressure sufficient to inject into the RCS. Alternatively, the boron storage tank may connect into the RPV 12 via pressure vessel feed-throughs separate from the ASF system. With continuing reference to FIGS. 1 and 2, the ASF system provides long-term cooling capability in a loss of heatsinking event. However, if in an operation 106 a safety threshold is exceeded (e.g., the pressure in the pressurizer 16 rises above a threshold pressure in spite of exhausting via the discharge line 50) then in an operation 110 the emergency core cooling system (ECCS) is suitably activated (for example, remediating the elevated pressure by venting into the radiological containment structure). The preferred embodiments have been illustrated and described. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof.
	summary
	abstract	A method for testing whether fuel rods of fuel assemblies resting on a working base and under water, of a nuclear reactor are leaking is disclosed. The method includes heating at least one first fuel assembly of a first division of fuel assemblies for driving radioactive fission products out of a defective fuel rod contained in the first fuel assembly. The first fuel assembly is continuously tested by extracting samples of water and continuously degassing the water removed from an area around the first fuel assembly even during the heating resulting in gas. A radioactivity of gaseous fission products released in the gas is continuously recorded. A fuel assembly belonging to a second division of fuel assemblies is heated only if the first fuel assembly belonging to the first division of fuel assemblies has been tested. An apparatus for implementing the method is also disclosed.
	abstract	Methods, apparatuses, devices, and systems for creating, controlling, conducting, and optimizing fusion activities of nuclei. The controlled fusion activities cover a spectrum of reactions from aneutronic, fusion reactions that produce essentially no neutrons, to neutronic, fusion reactions that produce substantial numbers of neutrons.
06160867&	description	DETAILED DESCRIPTION An X-ray-reflecting mirror according to a preferred embodiment of the invention comprises, on a suitable rigid substrate, layers of a first material and layers of a second material in alternating sequence in superposed fashion. Most preferably, each layer of the first material consists essentially of molybdenum (Mo), and each layer of the second material consists essentially of silicon (Si) as the principal constituent, with a dopant (preferably boron (B)) diffused into the silicon. By adjusting the dopant concentration in the silicon, the internal stress of the multi-layer reflective mirror can be controlled and minimized, compared to conventional multi-layer X-ray-reflecting mirrors, without having to rely on altering fabrication conditions and without reducing the reflectance of the mirror. The manner in which internal stress is understood to be controlled is as follows. When atoms diffuse into a crystalline substance, such diffusion can be of two types: "replacement diffusion" in which the diffusing atoms replace atoms of the substance at any of various lattice-point positions of the crystal lattice of the substance; and "penetration diffusion" in which the diffusing atoms penetrate between crystal-lattice positions of the substance. The type of substance into which the diffusion is occurring determines which will occur. For example, boron and certain other elements undergo replacement diffusion in silicon. The radius of a silicon atom is 1.17 .ANG., while the radius of a boron atom, as a representative dopant atom, is 0.88 .ANG.. Gilifalco, Atomic Diffusion in Crystals, Kyoritsu Publishing Co., 1980. I.e., the radius of a boron atom is only 75% the radius of a silicon atom; if a boron atom is inserted into a lattice point in a silicon crystal, the surrounding silicon atoms experience a tensile stress. As a result, the silicon crystal in general acquires a tensile stress. Tensile stress generated in this fashion increases as the boron concentration in the silicon is increased. For example, if 10.sup.20 atoms/cm.sup.3 of boron (approximately 0.1 atomic %) are introduced into monocrystalline Si, a tensile stress of about 100 MPa can be generated in the crystal. In a Mo/Si multi-layer mirror structure made using a sputtering technique, the molybdenum layers and the silicon layers have a compression stress due to the "peening" effect discussed above. The degree of such internal stress throughout the Mo/Si multi-layer mirror structure is dependent upon the conditions under which the layers were formed; in general, such stress is in the range of several tens of MPa to several hundreds of MPa. A representative multi-layer mirror structure according to the invention is shown in FIG. 1. A rigid substrate preferably made of glass or, alternatively, synthetic quartz ((e.g., "Zerodur" made by Schott or "ULE" made by Corning), or SiC, is provided with a mirror-polished surface. The substrate thickness is not critical so long as it has sufficient mechanical rigidity, stiffness, or stability. For example, a silicon substrate can have a thickness of 0.5 mm. Greater precision may require greater thickness, e.g., a thickness of 1/2, 1/3, or 1/4 the diameter of the substrate. In one representative embodiment, a layer 2 of a first material, preferably consisting essentially of molybdenum (Mo) (Mo is especially preferred for .lambda.=13 nm), is applied to the mirror-polished surface. As an alternative to Mo, the first material can consist essentially of Rh, Ru, Re, W, Ta, Ni, Cr, or Al, or any of various alloys of these materials. A layer 3 of a second material is applied superposedly to the first layer 2. The second material consists essentially of silicon (Si), as a principal ingredient, and a dopant. The dopant is preferably boron, but carbon or phosphorus can be used alternatively as the dopant. As an alternative to starting with the layer 2 of the first material on the substrate surface, it is possible to start with the layer 3 of the second material. In any event, additional first and second layers 2, 3 are superposedly applied in alternating sequence until the desired number of layers of the resulting multi-layer mirror structure has been formed. The number of layers is preferably in the range of 30-100, and most preferably about 50. Each of the layers 2, 3 is typically applied using any of various sputtering techniques, in which the "peening" effect becomes manifest (the peening effect generates a compression stress in the resulting structure). As an alternative to sputtering, vacuum evaporation can be used. The dopant is introduced into each of the silicon layers 3 by replacement diffusion (replacement diffusion of, e.g. , boron into crystalline silicon generates a tensile stress in the resulting structure). The tensile stress effectively offsets the compression stress imparted by sputtering to produce a substantially lower net stress in the structure. The tensile stress of the silicon layers increases as the dopant concentration in such layers increases. As a result, the net internal stress (i.e., the sum of the compression stress and tensile stress) of the multi-layer structure can be readily controlled and manipulated simply by adjusting the dopant concentration. FIG. 2 is a representative plot illustrating how the net internal stress in a Mo/Si multi-layer mirror structure (in this example, made with alternating layers of molybdenum as the first material and B-doped silicon as the second material applied by sputtering) can be reduced to zero. In FIG. 2, the abscissa is boron concentration in the silicon layers and the ordinate is net internal stress of the structure. Each molybdenum layer has substantially the same compression stress because, preferably, each molybdenum layer has the same thickness and was preferably applied using the same technique (e.g., sputtering) under similar conditions. The internal stress of the silicon layers changes from compression stress to tensile stress as the boron concentration increases in the silicon layers. At point A in FIG. 2, the total compression stress of the molybdenum layers and the total tensile stress of the silicon layers are equal and thus cancel each other in the overall structure. Thus, the net internal stress throughout the multi-layer structure is zero. In a multi-layer mirror structure according to this invention, the dopant concentration in the silicon layers is preferably 1.times.10.sup.18 atoms/cm.sup.3 (equal to 0.002 atomic %) or higher so as to provide a substantial reduction of the compression internal stress of the structure. The concentration of dopant is preferably the same in all the silicon layers. The thickness of each layer of the first material is not necessarily equal to the thickness of each layer of the second material, but the thickness of each layer of the first material is preferably the same. In any event, when the product of the "internal stress" and the "layer thickness" for the layers of the first material is equal to such a product for the layers of the second material, the internal stress of the multi-layer structure is zero. Introduction of dopant into the silicon (e.g., doping the silicon with boron) causes a change in the complex refractive index of the silicon. According to conventional wisdom, such a change would be expected to produce a significant decline in reflectance. However, as exemplified in FIG. 3, the introduction of boron into the silicon layers of the multi-layer structure causes substantially no loss of reflectance. Specifically, FIG. 3 depicts a representative reflectance (when the incident X-ray has a zero angle of incidence) from a Mo/Si multi-layer mirror structure comprising 50 layer pairs. Each layer of the first material in the FIG. 3 example has a thickness of 2.2 nm and each layer of the second material has a thickness of 4.5 nm. In the case of normal incidence, the condition for high reflectance is 2d =.lambda.. The wavelength deviation .DELTA..lambda. from .lambda. at which maximal reflectance is obtained is determined from the deviation .DELTA.d of the thickness d of a layer according to the following: .DELTA..lambda./.lambda.=.DELTA.d/d =.DELTA.d(Si)/d(Si)=.DELTA.d(Mo)/d(Mo). .DELTA.d(Si) and .DELTA.d(Mo) can be determined from the magnitude of deviation of the wavelength at which maximum reflectance is obtained, and from the magnitude of the change of reflectance at the wavelength .lambda. that is tolerable. FIG. 3 depicts curves obtained in the following four examples: (a) a multi-layer structure in which the silicon layers contain no boron, (b) a multi-layer structure in which the silicon layers contain 0.1 atomic % boron, (c) a multi-layer layer structure in which the silicon layers contain 1.0 atomic % boron, and (d) a multi-layer structure in which the silicon layers contain 10 atomic % boron. Peak wavelengths and reflectances for each of these four examples are listed below. ______________________________________ Boron Concentration Peak Wavelength Reflectance ______________________________________ 0 at % 131.6 .ANG. 74.3% 0.1 at % 131.6 .ANG. 74.3% 1 at % 131.6 .ANG. 74.2% 10 at % 131.4 .ANG. 73.4% ______________________________________ Hence, introduction of boron (as a representative dopant) into the silicon layers of a multi-layer mirror structure according to the invention causes essentially no decline in reflectance. As an alternative to molybdenum as a first material, any of various other materials can be used such as Rh, Re, W, Ta, Ni, Cr, Al, and alloys of such materials. The invention will be better understood by reference to the following example embodiments, which are intended to merely illustrate the best mode now known for practicing the invention. The scope of the invention is not to be considered limited to these example embodiments. EXAMPLE EMBODIMENT 1 A multi-layer mirror structure according to this example embodiment was formed using an ion-beam sputtering technique. Specifically, for forming the layers of the first material and the layers of the second material, a molybdenum (Mo) sputtering target and a silicon (Si) sputtering target, respectively, were used. The Si target had a boron (B) concentration of 0.5 atomic %. The resulting multi-layer mirror structure was formed by alternately layering the Mo layers 2 and the B-diffused Si layers 3 onto a mirror-polished surface of a glass substrate 1 (FIG. 1). The period length of the multi-layer mirror structure was 6.7 nm (the thickness of each Mo layer was 4.5 nm and the thickness of each Si layer was 2.2 nm). Fifty pairs of layers were applied to the substrate. For comparison, a Mo target and a Si target containing no boron were used to make, using an ion-beam sputtering technique, a "control" multi-layer mirror structure having the same period length, layer thicknesses, and number of layers as the example embodiment. The net internal stresses of the example embodiment and of the control were calculated, using conventional methods, from measured substrate warping (bending) before and after the formation of the fifty pairs of layers. The calculations can be performed in at least two ways: (a) sum up all the Fresnel coefficients of each interlayer boundary as an interference effect; or (b) perform a matrix analysis as described in Born and Wolf, Principles of Optics, 5.sup.th edition, pp. 51-70, Pergamon Press, 1975, incorporated herein by reference. The net internal stress of the control was approximately 300 MPa of compression stress. The net internal stress of the example embodiment was approximately 10 MPa of tensile stress. This example embodiment exhibited reflectance behavior as profiled in FIG. 3. Thus, this example embodiment exhibited a substantial reduction of internal stress without a decline in reflectance. EXAMPLE EMBODIMENT 2 A multi-layer mirror structure according to this example embodiment was made using a high-frequency magnetron sputtering technique. To make each layer of the first material, a Mo sputtering target was used; to make each layer of the second material, a Si sputtering target having a boron (B) concentration of 0.4 atomic % was used. The Mo layers 2 and the B-doped Si layers 3 were alternately applied to a mirror-polished surface of a glass substrate 1 (FIG. 1). The period length was 6.7 nm (the thickness of each Mo layer was 4.5 nm and the thickness of each Si layer was 2.2 nm), and 50 pairs of layers were applied. As a control, an otherwise similar multi-layer mirror structure was formed using the same technique but in which the Si layers contained no boron. The control had the same periodic length, layer thicknesses, and number of layers as the example embodiment. The net internal stresses of each multi-layer mirror structure were calculated from measured warping (bending) of the substrate before and after the layers were applied. The net internal stress of the control was approximately 300 MPa of compression stress, and the net internal stress of the example embodiment was approximately 5 MPa of tensile stress. This example embodiment exhibited a reflectance profile as shown in FIG. 3. Thus, this example embodiment exhibited a substantially reduced net internal stress without a significant decline in reflectance. EXAMPLE EMBODIMENT 3 A multi-layer mirror structure according to this example embodiment was formed using an ion-beam sputtering technique. To make each layer of the first material, sputtering was performed using a Mo sputtering target. To make each layer of the second material, sputtering was performed using a compound sputtering target. The compound sputtering target comprised Si extending over a portion of a boron target surface, wherein the surface-area ratio of boron (B) to Si on the target surface was 1:5. The multi-layer mirror structure was formed by alternately layering Mo layers 2 and B-doped Si layers 3 on a mirror-polished surface of a glass substrate 1 (FIG. 1). The period length of the multi-layer mirror structure was 6.7 nm (the thickness of each molybdenum layer was 4.5 nm and the thickness of each silicon layer was 2.2 nm). Fifty pairs of layers were applied. For comparison, a "control" structure comprising alternating layers of Mo and Si (without any boron) was formed using the same sputtering technique. The control had the same period length, layer thicknesses and number of layers as the example embodiment. The net internal stresses of the example embodiment and of the control were determined from measurements of warping (bending) of the substrate before and after forming the layers on the substrate. The net internal stress of the control was approximately 300 MPa of compression stress, and the net internal stress of the example embodiment was approximately 10 MPa of tensile stress. This example embodiment exhibited a reflectance profile as shown in FIG. 3. Thus, this example embodiment exhibited a substantial reduction of net internal stress without a significant decline in reflectance. EXAMPLE EMBODIMENT 4 A multi-layer mirror structure according to this example embodiment was made using a high-frequency magnetron sputtering technique. To make each layer of the first material, sputtering using a Mo sputtering target was performed. To make each layer of the second material, sputtering using a compound sputtering target was performed. The compound sputtering target comprised a Si wafer extending over a portion of a boron (B) target surface, wherein the ratio of B-to-Si surface area on the target surface was 1:6. The multi-layer mirror structure was formed by alternately layering Mo layers 2 and B-doped Si layers 3 onto a mirror-polished surface of a glass substrate 1 (FIG. 1). The period length of this example embodiment was 6.7 nm (the thickness of each Mo layer was 4.5 nm, and the thickness of each Si layer was 2.2 nm). Fifty pairs of layers were formed. For comparison, a "control" structure was formed using the same sputtering technique, except that the Si layers contained no boron. The control had the same period length, layer thicknesses, and number of layers as the example embodiment. The net internal stresses of the control and of the example embodiment were calculated from measurements of warping (bending) of the substrate before and after application of the layers. The net internal stress of the control was approximately 300 MPa of compression stress, but the net internal stress of the example embodiment was approximately 5 MPa of tensile stress. This example embodiment exhibited a reflectance profile as shown in FIG. 3. Thus, this example embodiment exhibited a substantially reduced internal stress without a significant decline in reflectance. Each of the example embodiments exhibited a reduced net internal stress, compared to controls lacking dopant in the silicon layers, that was independent of the technique used to form the multi-layer mirror structure. The example embodiments also demonstrated that the net internal stress of a multi-layer mirror structure according to the present invention can be freely controlled by manipulating the concentration of dopant in the silicon layers. For example, an otherwise substantial net compression stress can be eliminated entirely or converted to a slight net tensile stress. This allows excellent control over warping of the structure due to excessive internal stress, with a consequent remarkable improvement in optical performance of the multi-layer mirror structure. Whereas the invention has been described in connection with a preferred embodiment and multiple example embodiments, it will be understood that the invention is not limited to those embodiments. On the contrary, the invention is intended to encompass all modifications, alternatives, and equivalents as may be encompassed within the spirit and scope of the invention as defined by the appended claims.
	claims	1. A device comprising: an underground vertical shaft having a top and a bottom; a nuclear reactor envelope suspended in said shaft at said top; a nuclear reactor in said envelope; an activatable release mechanism to release said nuclear reactor envelope when desired; whereby activating said release mechanism causes said nuclear reactor envelope and said nuclear reactor to fall down to said bottom of said shaft further comprising a drainage pipe having an upper end and a lower end, said upper end being accessible from ground level and said lower end opening into a lower portion of said vertical shaft; flowable material located in said drainage pipe and adjacent to said upper end; an activatable gate abutting against said lower end, operably connected to said nuclear reactor, whereby falling of said nuclear reactor past said lower end activates said gate and causes said flowable material to flow through said drainage pipe into said vertical shaft on top of said nuclear reactor. 2. A device according to claim 1 , further comprising: claim 1 a supply pipe having an upper portion accessible from ground level and a lower portion opening into said vertical shaft below said reactor, whereby filler placed in said supply pipe falls out of said lower portion and into said vertical shaft; a pulverizing device positioned below said lower portion for pulverizing spent nuclear fuel to form pulverized nuclear fuel, positioned in said shaft below said lower portion of said supply pipe, whereby said pulverized fuel mixes with said filler and falls to said bottom of said shaft. 3. A device according to claim 2 , wherein said pulverizing device mixes said pulverized nuclear fuel with said filler to form a mixture and places said mixture in a bag. claim 2 4. A device according to claim 2 , wherein said pulverizing device is selected from the group consisting of a sand blasting machine and a grinder. claim 2 5. A device, comprising: a vertical shaft approximately 1200 feet deep and approximately 60 feet in diameter having a top and a bottom; a support structure at the top of the shaft; at least one stem extending downwardly from the support structure; a nuclear reactor envelope suspended from the stem over the shaft; a nuclear reactor enclosed in the envelope; a release mechanism operably connected to the stem, whereby activating said release mechanism severs the stem and causes the reactor and the envelope to fall to the bottom of the shaft; a drainage pipe having an upper end and a lower end, the upper end being accessible from ground level and the lower end opening into a lower portion of said vertical shaft; a hopper attached to the upper end; flowable material located in the drainage pipe and the hopper; an activatable gate abutting against the lower end operably connected to the nuclear reactor envelope, whereby falling of the nuclear reactor envelope past the lower end activates the gate and causes the flowable material to flow through the drainage pipe into the vertical shaft and to fall on top of the nuclear reactor envelope and the nuclear reactor; a supply pipe having an upper portion accessible from ground level and a lower portion which opens into the vertical shaft, whereby filler placed into the supply pipe falls out of the lower portion and into the vertical shaft; a pulverizing device positioned below the lower portion for pulverizing spent nuclear fuel to form pulverized nuclear fuel, positioned near the shaft below the nuclear reactor envelope, whereby the pulverized nuclear fuel and filler can fall into the shaft; wherein the pulverizing device mixes the pulverized nuclear fuel with the filler to form a mixture and places the mixture in a bag; wherein the pulverizing device is selected from the group consisting of a sand blasting machine and a grinder. 6. A device according to claim 5 , wherein the flowable material is dry sand. claim 5 7. A device according to claim 6 , wherein surfaces of the gate, the hopper and the drainage pipe in contact with the flowable material are coated with waterproof sealant. claim 6 8. A device according to claim 5 , wherein the release mechanism is selected from the group comprising an electric furnace and explosives. claim 5 9. A device according to claim 5 , further comprising additional stems between the nuclear reactor envelope and the support structure. claim 5
048308161	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS The getter trap 1 (FIG. 1) of the present invention includes an elongated, closed housing 3 having an inlet 5 at one end thereof and an outlet 7 at the other end. A getter material 9 (FIG. 2) is randomly disposed within the housing 3. The getter material 9 is in the form of hollow, tubular sections 11, which are held in place within the housing 3 by a pair of baffle plates 13. The tubular sections 11 have a zirconium-containing substrate 15, formed of a material such as Zircaloy-2, which is an alloy 1.2 to 1.7 percent tin, 0.07-0.20 percent iron, 0.05-0.15 percent chromium, 0.03-0.08 percent nickel, and the balance zirconium; or Zircaloy-4, which is an alloy 1.2 to 1.7 percent tin, 0.12-0.18 percent iron, 0.05-0.15 percent chromium, and the balance zirconium. Zircaloy-4 tubes having a length of 1 inch, and outer diameter of 0.375 inch, and a wall thickness of 0.023 inch are particularly useful in the getter material 9 for the getter trap 1 of the invention. The hollow, tubular sections 11 are coated with a coating 17 of the gettering alloy disclosed in U.S. Pat. No. 4,312,669 of zirconium, vanadium and iron, whose composition, in weight percent, when plotted on a ternary diagram (FIG. 3), lies within a triangle having as its corners the points defined by: (a) 75 percent zirconium, 20 percent vanadium and 5 percent iron; PA1 (b) 45 percent zirconium, 20 percent vanadium and 35 percent iron; and PA1 (c) 45 percent zirconium, 50 percent vanadium and 5 percent iron. Preferably, the composition of the gettering alloy, when plotted on the ternary diagram of FIG. 3 lies within a polygon having as its corners the points defined by: PA1 (d) 70 percent zirconium, 25 percent vanadium and 5 percent iron; PA1 (e) 70 percent zirconium, 24 percent vanadium and 6 percent iron; and PA1 (f) 66 percent zirconium, 24 percent vanadium and 10 percent iron. PA1 (g) 47 percent zirconium, 43 percent vanadium and 10 percent iron; PA1 (h) 47 percent zirconium, 45 percent vanadium and 8 percent iron; and PA1 (i) 50 percent zirconium, 45 percent vanadium and 5 percent iron. One particularly suitably gettering alloy has a composition of 45 percent zirconium, 50 percent vanadium and 5 percent iron. Tubes of Zircaloy-4 coated with this gettering alloy are available from SAES Getters S.p.A. as SAES alloy type ST123. The gettering alloy has a porous structure, and consequently, has a high surface area and a high rate of absorption for both oxygen and hydrogen. In addition, the gettering alloy has a high coefficient of diffusion for both hydrogen and oxygen. Thus, the oxygen absorbed by the gettering alloy diffuses from the surface of the getter material 9 into the bulk of the getter material 9 instead of forming a hydrogen absorption-inhibiting oxide film on the surface of the getter material 9. Access by hydrogen to the zirconium-containing substrate when not coated with the gettering alloy is normally greatly inhibited by the formation of tightly adherent chemical films of the oxide and nitride of zirconium. The coating 17 of the gettering alloy on the surface of the getter material 9 allows the hydrogen to gain access to the zirconium-containing substrate 15. This provides a higher density of hydrogen loading than would the gettering alloy alone, because of the higher density of zirconium in the zirconium-containing substrate than in the porous gettering alloy, and because of the higher chemical affinity of the zirconium-containing substrate for hydrogen in comparison with the gettering alloy. The chemical thermodynamic properties of the zirconium-containing substrate predominate, and at high hydrogen loading, these properties are more advantageous than those of the gettering alloy. For instance, the getter material of the invention will operate at a temperature that is typical of the secondary liquid metal coolant in the secondary coolant loop of a liquid metal nuclear reactor after the secondary coolant has been cooled y passage through the second heat exchanger. As a liquid metal, such as liquid sodium, flows through the inlet 5 into the housing 3 and through the getter material 9, and is discharged from the housing 3 through the outlet 7, hydrogen and oxygen impurities are removed from the liquid metal. The distribution of absorbed hydrogen and oxygen decreases exponentially from the inlet 5 along the length of the getter trap 1. If sufficiently large amounts of oxygen are absorbed, then a proportion of the upstream getter material 9 could have reduced effectiveness and capacity for simultaneous hydrogen absorption. Thus, these considerations can be taken into account in designing the getter trap 1, with an upstream section of getter material 9 being in effect sacrificed for oxygen removal. The absorption of hydrogen by the gettering alloy is reversible by heating the getter material 9 to a temperature on the order of 700.degree. C. The absorption of oxygen by the gettering alloy is not reversible. Thus, the design of a getter trap 1 must take into consideration the life expectancy of the getter material 9 due to oxygen absorption, based upon the environment in which the particular getter trap 1 is to be used. The getter trap 1 of the invention can include a second getter material 21 comprising pellets 27 of the gettering alloy. Such a second getter material 21 is available from SAES Getters S.p.A. as SAES alloy type ST172. Thus, in a second embodiment of the invention, illustrated in FIG. 4, the housing 3 of the getter trap 1A includes a first section 19 in which a second getter material 21 is randomly disposed. The getter material 9 is randomly disposed in a second section 23, located downstream of the first section 19. The first section 19 and second section 23 of the housing 3 are preferably separated by means, such as a baffle plate 25. The upstream, second getter material 21 operates primarily as an oxygen getter, enabling the getter material 9, located downstream of the second getter material 21, to operate primarily as a hydrogen getter without interference from oxygen being absorbed by the getter material 9. Alternatively, in a third embodiment of the invention, illustrated in FIG. 5, in getter trap 1B, the second getter material 21 is randomly disposed in a second housing 29, located directly upstream of the housing 3 in which the getter material 9 is randomly disposed. The second housing 29 includes an inlet 31 and an outlet 33. The second getter material 21 is held in place within the second housing 29 by a pair of baffle plates 35. A connection 37 secures the outlet 33 of the second housing 29 and the inlet 5 of the first housing 3 togeter. Thus, the liquid metal containing hydrogen and oxygen impurities flows into the getter trap 1B through the inlet 31 of the second housing 29, flows through the second getter material 21, and out of the second housing 29 through the outlet 33. The liquid metal then flows through the inlet 5, flows through the getter material 9, and out of the getter trap 1B through the outlet 7 of the housing 3. These embodiments of the getter trap are expected to be particularly useful in the secondary coolant loop of a liquid metal cooled nuclear reactor, where the oxygen load is small compared to hydrogen load arising from corrosion of the steam-containing third loop. The getter trap 1 of the invention is particularly useful in a liquid metal cooled nuclear reactor system 39 (FIG. 6). The system 39 includes a primary coolant loop 41 wherein the primary coolant, a liquid metal such as liquid sodium, is heated as it passes through the nuclear core of the liquid metal nuclear reactor 43. The primary liquid metal coolant is then cooled by passing through a first heat exchanger 45, and is directed back to the reactor 43 by a pump 47. Liquid metal coolant, circulating in a secondary coolant loop 49, absorbs thermal energy from the primary liquid metal coolant through indirect heat exchange in the first heat exchanger 45. The heated secondary liquid metal coolant then passes through a second heat exchanger 51 wherein the thermal energy from the secondary liquid metal coolant is used to heat water flowing in a third loop 53 to produce steam, which, in turn, is used to drive a steam driven device (not shown), such as an electrical generator. The secondary liquid metal coolant is then directed from the second heat exchanger 51 back to the first heat exchanger 45 by a pump 55. The secondary liquid metal coolant often contains undesired hydrogen from the water used to operate the steam driven device, and oxygen from other sources. Thus, the secondary coolant loop 49 includes a getter trap 1, preferably in the embodiments illustrated in FIGS. 4 and 5, of the invention downstream of the second heat exchanger 51 to eliminate the undesired hydrogen and oxygen from the secondary liquid metal coolant. Flow to the getter trap 1 is controlled by a valve 57. It is easily understood that the getter trap of the invention can be used in the primary coolant loop in situations where the primary coolant includes hydrogen and oxygen contaminants. In the primary coolant loop, containing undesired hydrogen and oxygen, the getter trap 1' would be included, following a valve 57', preferably in the line between the pump 47 and the liquid metal reactor 43. The insensitivity of the getter trap to the temperature of operation would also, alternatively, allow the same to be placed in the hot inlet line to the heat exchanger 45 of the primary coolant loop 41, or in the hot inlet line to the heat exchanger 51 of the secondary coolant loop 49.
	description	The present application is related to and claims the benefit of the earliest available effective filing date(s) from the following listed application(s) (the “Related Applications”) (e.g., claims earliest available priority dates for other than provisional patent applications or claims benefits under 35 USC §119(e) for provisional patent applications, for any and all parent, grandparent, great-grandparent, etc. applications of the Related Application(s)). All subject matter of the Related Applications and of any and all parent, grandparent, great-grandparent, etc. applications of the Related Applications is incorporated herein by reference to the extent such subject matter is not inconsistent herewith. For purposes of the USPTO extra-statutory requirements, the present application constitutes a continuation-in-part of U.S. patent application Ser. No. 11/605,933, entitled CONTROLLABLE LONG TERM OPERATION OF A NUCLEAR REACTOR, naming Roderick A. Hyde, Muriel Y. Ishikawa, Nathan P. Myhrvold, and Lowell L. Wood, Jr., as inventors, filed 28 Nov. 2006, which is currently co-pending, or is an application of which a currently co-pending application is entitled to the benefit of the filing date. For purposes of the USPTO extra-statutory requirements, the present application is related to U.S. patent application Ser. No. 11/605,943, entitled AUTOMATED NUCLEAR POWER REACTOR FOR LONG-TERM OPERATION, naming Roderick A. Hyde, Muriel Y. Ishikawa, Nathan P. Myhrvold, and Lowell L. Wood, Jr., as inventors, filed 28 Nov. 2006, which is currently co-pending, or is an application of which a currently co-pending application is entitled to the benefit of the filing date. For purposes of the USPTO extra-statutory requirements, the present application is related to U.S. patent application Ser. No. 11/605,848, entitled METHOD AND SYSTEM FOR PROVIDING FUEL IN A NUCLEAR REACTOR, naming Roderick A. Hyde, Muriel Y. Ishikawa, Nathan P. Myhrvold, and Lowell L. Wood, Jr., as inventors, filed 28 Nov. 2006, which is currently co-pending, or is an application of which a currently co-pending application is entitled to the benefit of the filing date. The United States Patent Office (USPTO) has published a notice to the effect that the USPTO's computer programs require that patent applicants reference both a serial number and indicate whether an application is a continuation or continuation-in-part. Stephen G. Kunin, Benefit of Prior-Filed Application, USPTO Official Gazette Mar. 18, 2003, available at http://www.uspto.gov/web/offices/com/sol/og/2003/week11/patbene.htm. The present Applicant Entity (hereinafter “Applicant”) has provided above a specific reference to the application(s) from which priority is being claimed as recited by statute. Applicant understands that the statute is unambiguous in its specific reference language and does not require either a serial number or any characterization, such as “continuation” or “continuation-in-part,” for claiming priority to U.S. patent applications. Notwithstanding the foregoing, Applicant understands that the USPTO's computer programs have certain data entry requirements, and hence Applicant is designating the present application as a continuation-in-part of its parent applications as set forth above, but expressly points out that such designations are not to be construed in any way as any type of commentary and/or admission as to whether or not the present application contains any new matter in addition to the matter of its parent application(s). All subject matter of the Related Applications and of any and all parent, grandparent, great-grandparent, etc. applications of the Related Applications is incorporated herein by reference to the extent such subject matter is not inconsistent herewith. The present application relates to nuclear reactors, and systems, applications, and apparatuses related thereto. The following embodiments and aspects thereof are described and illustrated in conjunction with systems and methods which are meant to be exemplary and illustrative, not limiting in scope. Exemplary embodiments provide automated nuclear fission reactors and methods for their operation. Exemplary embodiments and aspects include, without limitation, re-use of nuclear fission fuel, alternate fuels and fuel geometries, modular fuel cores, fast fluid cooling, variable burn-up, programmable nuclear thermostats, fast flux irradiation, temperature-driven neutron absorption, low coolant temperature cores, refueling, and the like. In addition to the exemplary embodiments and aspects described above, further embodiments and aspects will become apparent by reference to the drawings and by study of the following detailed description. By way of overview, embodiments provide automated nuclear fission reactors and methods for their operation. Details of an exemplary reactor, exemplary core nucleonics, and operations, all given by way of non-limiting example, will be set forth first. Then, details will be set forth regarding several exemplary embodiments and aspects, such as without limitation re-use of nuclear fission fuel, alternate fuels and fuel geometries, modular fuel cores, fast fluid cooling, variable burn-up, programmable nuclear thermostats, fast flux irradiation, temperature-driven neutron absorption, low coolant temperature cores, refueling, and the like. Referring now to FIG. 1A, a nuclear fission reactor 10, given by way of example and not of limitation, acts as an exemplary host environment for embodiments and aspects described herein. While many embodiments of the reactor 10 are contemplated, a common feature among many contemplated embodiments of the reactor 10 is origination and propagation of a nuclear fission deflagration wave, or “burnfront”. Considerations Before discussing details of the reactor 10, some considerations behind embodiments of the reactor 10 will be given by way of overview but are not to be interpreted as limitations. Some embodiments of the reactor 10 reflect attainment of all of the considerations discussed below. On the other hand, some other embodiments of the reactor 10 reflect attainment of selected considerations, and need not accommodate all of the considerations discussed below. Portions of the following discussion includes information excerpted from a paper entitled “Completely Automated Nuclear Power Reactors For Long-Term Operation: III. Enabling Techology For Large-Scale, Low-Risk, Affordable Nuclear Electricity” by Edward Teller, Muriel Ishikawa, Lowell Wood, Roderick Hyde, and John Nuckolls, “PRESENTED AT the July 2003 Workshop of the Aspen Global Change institute”, University of California Lawrence Livermore National Laboratory publication UCRL-JRNL-122708 (2003). (This paper was prepared for submittal to Energy, The International Journal, 30 Nov. 2003.) the entire contents of which are hereby incorporated by reference. Nuclear fission fuels envisioned for use in embodiments of the reactor 10 are typically widely available, such as without limitation uranium (natural, depleted, or enriched), thorium, plutonium, or even previously-burned nuclear fission fuel assemblies. Other, less widely available nuclear fission fuels, such as without limitation other actinide elements or isotopes thereof may be used in embodiments of the reactor 10. While embodiments of the reactor 10 contemplate long-term operation at full power on the order of around ⅓ century to around ½ century or longer, an aspect of some embodiments of the reactor 10 does not contemplate nuclear refueling (but instead contemplate burial in-place at ends-of-life) while some aspects of embodiments of the reactor 10 contemplate nuclear refueling—with some nuclear refueling occurring during shutdown and some nuclear refueling occurring during operation at power. It is also contemplated that nuclear fission fuel reprocessing may be avoided, thereby mitigating possibilities for diversion to military uses and other issues. Other considerations behind embodiments of the reactor 10 include disposing in a manifestly safe manner long-lived radioactivity generated in the course of operation. It is envisioned that the reactor 10 may be able to mitigate damage due to operator error, casualties such as a loss of coolant accident (LOCA), or the like. In some aspects decommissioning may be effected in low-risk and inexpensive manner. As a result, some embodiments of the reactor 10 may entail underground siting, thereby addressing large, abrupt releases and small, steady-state releases of radioactivity into the biosphere. Some embodiments of the reactor 10 may entail minimizing operator controls, thereby automating those embodiments as much as practicable. In some embodiments, a life-cycle-oriented design is contemplated, wherein those embodiments of the reactor 10 can operate from startup to shutdown at end-of-life in as fully-automatic manner as practicable. Some embodiments of the reactor 10 lend themselves to modularized construction. Finally, some embodiments of the reactor 10 may be designed according to high power density. Some features of various embodiments of the reactor 10 result from some of the above considerations. For example, simultaneously accommodating desires to achieve ⅓-½ century (or longer) of operations at full power without nuclear refueling and to avoid nuclear fission fuel reprocessing entails use of a fast neutron spectrum. As another example, in some embodiments a negative temperature coefficient of reactivity (αT) is engineered-in to the reactor 10, such as via negative feedback on local reactivity implemented with strong absorbers of fast neutrons. As a further example, in some embodiments of the reactor 10 a distributed thermostat enables a propagating nuclear fission deflagration wave mode of nuclear fission fuel burn. This mode simultaneously permits a high average burn-up of non-enriched actinide fuels, such as natural uranium or thorium, and use of a comparatively small “nuclear fission igniter” region of moderate isotopic enrichment of nuclear fissionable materials in the core's fuel charge. As another example, in some embodiments of the reactor 10, multiple redundancy is provided in primary and secondary core cooling. Exemplary Embodiment of Nuclear Fission Reactor Now that some of the considerations behind some of the embodiments of the reactor 10 have been set forth, further details regarding an exemplary embodiment of the reactor 10 will be explained. It is emphasized that the following description of an exemplary embodiment of the reactor 10 is given by way of non-limiting example only and not by way of limitation. As mentioned above, several embodiments of the reactor 10 are contemplated, as well as further aspects of the reactor 10. After details regarding an exemplary embodiment of the reactor 10 are discussed, other embodiments and aspects will also be discussed. Still referring to FIG. 1A, an exemplary embodiment of the reactor 10 includes a nuclear fission reactor core assembly 100 that is disposed within a reactor pressure vessel 12. Several embodiments and aspects of the nuclear fission reactor core assembly 100 are contemplated that will be discussed later. Some of the features that will be discussed later in detail regarding the nuclear fission reactor core assembly 100 include nuclear fission fuel materials and their respective nucleonics, fuel assemblies, fuel geometries, and initiation and propagation of nuclear fission deflagration waves. The reactor pressure vessel 12 suitably is any acceptable pressure vessel known in the art and may be made from any materials acceptable for use in reactor pressure vessels, such as without limitation stainless steel. Within the reactor pressure vessel 12, a neutron reflector (not shown) and a radiation shield (not shown) surround the nuclear fission reactor core assembly 100. In some embodiments, the reactor pressure vessel 12 is sited underground. In such cases, the reactor pressure vessel 12 can also function as a burial cask for the nuclear fission reactor core assembly 100. In these embodiments, the reactor pressure vessel 12 suitably is surrounded by a region (not shown) of isolation material, such as dry sand, for long-term environmental isolation. The region (not shown) of isolation material may have a size of around 100 m in diameter or so. However, in other embodiments, the reactor pressure vessel 12 is sited on or toward the Earth's surface. Reactor coolant loops 14 transfer heat from nuclear fission in the nuclear fission reactor core assembly 100 to application heat exchangers 16. The reactor coolant may be selected as desired for a particular application. In some embodiments, the reactor coolant suitably is helium (He) gas. In other embodiments, the reactor coolant suitably may be other pressurized inert gases, such as neon, argon, krypton, xenon, or other fluids such as water or gaseous or superfluidic carbon dioxide, or liquid metals, such as sodium or lead, or metal alloys, such as Pb—Bi, or organic coolants, such as polyphenyls, or fluorocarbons. The reactor coolant loops suitably may be made from tantalum (Ta), tungsten (W), aluminum (Al), steel or other ferrous or non-iron groups alloys or titanium or zirconium-based alloys, or from other metals and alloys, or from other structural materials or composites, as desired. In some embodiments, the application heat exchangers 16 may be steam generators that generate steam that is provided as a prime mover for rotating machinery, such as electrical turbine-generators 18 within an electrical generating station 20. In such a case, the nuclear fission reactor core assembly 100 suitably operates at a high operating pressure and temperature, such as above 1,000K or so and the steam generated in the steam generator may be superheated steam. In other embodiments, the application heat exchanger 16 may be any steam generator that generates steam at lower pressures and temperatures (that is, need not be not superheated steam) and the nuclear fission reactor core assembly 100 operates at temperatures less than around 550K. In these cases, the application heat exchangers 16 may provide process heat for applications such as desalination plants for seawater or for processing biomass by distillation into ethanol, or the like. Optional reactor coolant pumps 22 circulate reactor coolant through the nuclear fission reactor core assembly 100 and the application heat exchangers 16. Note that although the illustrative embodiment shows pumps and gravitationally driven circulation, other approaches may not utilize pumps, or circulatory structures or be otherwise similarly geometrically limited. The reactor coolant pumps 22 suitably are provided when the nuclear fission reactor core assembly 100 is sited approximately vertically coplanar with the application heat exchangers 16, such that thermal driving head is not generated. The reactor coolant pumps 22 may also be provided when the nuclear fission reactor core assembly 100 is sited underground. However, when the nuclear fission reactor core assembly 100 is sited underground or in any fashion so the nuclear fission reactor core assembly 100 is vertically spaced below the application heat exchangers 16, thermal driving head may be developed between the reactor coolant exiting the reactor pressure vessel 12 and the reactor coolant exiting the application heat exchangers 16 at a lower temperature than the reactor coolant exiting the reactor pressure vessel 12. When sufficient thermal driving head exists, the reactor coolant pumps 22 need not be provided in order to provide sufficient circulation of reactor coolant through the nuclear fission reactor core assembly 100 to remove heat from fission during operation at power. In some embodiments more than one reactor coolant loop 14 may be provided, thereby providing redundancy in the event of a casualty, such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA) or a primary-to-secondary leak or the like, to any one of the other reactor coolant loops 14. Each reactor coolant loop 14 is typically rated for full-power operation, though some applications may remove this constraint. In some embodiments, one-time closures 24, such as reactor coolant shutoff valves, are provided in lines of the reactor coolant system 14. In each reactor coolant loop 14 provided, a closure 24 is provided in an outlet line from the reactor pressure vessel 12 and in a return line to the reactor pressure vessel 12 from an outlet of the application heat exchanger 16. The one-time closures 24 are fast-acting closures that shut quickly under emergency conditions, such as detection of significant fission-product entrainment in reactor coolant). The one-time closures 24 are provided in addition to a redundant system of automatically-actuated conventional valves (not shown). Heat-dump heat exchangers 26 are provided for removal of after-life heat (decay heat). The heat-dump heat exchanger 26 includes a primary loop that is configured to circulate decay heat removal coolant through the nuclear fission reactor core assembly 100. The heat-dump heat exchanger 26 includes a secondary loop that is coupled to an engineered heat-dump heat pipe network (not shown). In some situations, for example, for redundancy purposes, more than one the heat-dump heat exchanger 26 may be provided. Each of the heat-dump heat exchangers 26 provided may be sited at a vertical distance above the nuclear fission reactor core assembly 100 so sufficient thermal driving head is provided to enable natural flow of decay heat removal coolant without need for decay heat removal coolant pumps. However, in some embodiments decay heat removal pumps (not shown) may be provided or, if provided, the reactor coolant pumps may be used for decay heat removal, where appropriate. Now that an overview of an exemplary embodiment of the reactor 10 has been given, other embodiments and aspects will be discussed. First, embodiments and aspects of the nuclear fission reactor core assembly 100 will be discussed. An overview of the nuclear fission reactor core assembly 100 and its nucleonics and propagation of a nuclear fission deflagration wave will be set forth first, followed by descriptions of exemplary embodiments and other aspects of the nuclear fission reactor core assembly 100. Given by way of overview and in general terms, structural components of the reactor core assembly 100 may be made of tantalum (Ta), tungsten (W), rhenium (Re), or carbon composite, ceramics, or the like. These materials are suitable because of the high temperatures at which the nuclear fission reactor core assembly 100 operates, and because of their creep resistance over the envisioned lifetime of full power operation, mechanical workability, and corrosion resistance. Structural components can be made from single materials, or from combinations of materials (e.g., coatings, alloys, multilayers, composites, and the like). In some embodiments, the reactor core assembly 100 operates at sufficiently lower temperatures so that other materials, such as aluminum (Al), steel, titanium (Ti) or the like can be used, alone or in combinations, for structural components. The nuclear fission reactor core assembly 100 includes a small nuclear fission igniter and a larger nuclear fission deflagration burn-wave-propagating region. The nuclear fission deflagration burn-wave-propagating region suitably contains thorium or uranium fuel, and functions on the general principle of fast neutron spectrum fission breeding. In some embodiments, uniform temperature throughout the nuclear fission reactor core assembly 100 is maintained by thermostating modules, described in detail later, which regulate local neutron flux and thereby control local power production. The nuclear fission reactor core assembly 100 suitably is a breeder for reasons of efficient nuclear fission fuel utilization and of minimization of requirements for isotopic enrichment. Further, and referring now to FIGS. 1B and 1C, the nuclear fission reactor core assembly 100 suitably utilizes a fast neutron spectrum because the high absorption cross-section of fission products for thermal neutrons does not permit utilization of more than about 1% of thorium or of the more abundant uranium isotope, U238, in uranium-fueled embodiments, without removal of fission products. In FIG. 1B, cross-sections for the dominant neutron-driven nuclear reactions of interest for the Th232-fueled embodiments are plotted over the neutron energy range 10−3-107 eV. It can be seen that losses to radiative capture on fission product nuclei dominate neutron economies at near-thermal (˜0.1 eV) energies, but are comparatively negligible above the resonance capture region (between ˜3-300 eV). Thus, operating with a fast neutron spectrum when attempting to realize a high-gain fertile-to-fissile breeder can help to preclude fuel recycling (that is, periodic or continuous removal of fission products). The radiative capture cross-sections for fission products shown are those for intermediate-Z nuclei resulting from fast neutron-induced fission that have undergone subsequent beta-decay to negligible extents. Those in the central portions of the burn-waves of embodiments of the nuclear fission reactor core assembly 100 will have undergone some decay and thus will have somewhat higher neutron avidity. However, parameter studies have indicated that core fuel-burning results may be insensitive to the precise degree of such decay. In FIG. 1C, cross-sections for the dominant neutron-driven nuclear reactions of primary interest for the Th232-fueled embodiments are plotted over the most interesting portion of the neutron energy range, between >104 and <106.5 eV, in the upper portion of FIG. 1C. The neutron spectrum of embodiments of the reactor 10 peaks in the ≧105 eV neutron energy region. The lower portion of FIG. 1C contains the ratio of these cross-sections vs. neutron energy to the cross-section for neutron radiative capture on Th232, the fertile-to-fissile breeding step (as the resulting Th233 swiftly beta-decays to Pa233, which then relatively slowly beta-decays to U233, analogously to the U239—Np239—Pu239 beta decay-chain upon neutron capture by U238). It can be seen that losses to radiative capture on fission products are comparatively negligible over the neutron energy range of interest, and furthermore that atom-fractions of a few tens of percent of high-performance structural material, such as Ta, will impose tolerable loads on the neutron economy in the nuclear fission reactor core assembly 100. These data also suggest that core-averaged fuel burn-up in excess of 50% can be realizable, and that fission product-to-fissile atom-ratios behind the nuclear fission deflagration wave when reactivity is finally driven negative by fission-product accumulation will be approximately 10:1. Origination and Propagation of Nuclear Fission Deflagration Wave Burnfront The nuclear fission deflagration wave within the nuclear fission reactor core assembly 100 will now be explained. Propagation of deflagration burning-waves through combustible materials can release power at a predictable level. Moreover, if the material configuration has the requisite time-invariant features, the ensuing power production may be at a steady level. Finally, if deflagration wave propagation-speed may be externally modulated in a practical manner, the energy release-rate and thus power production may be controlled as desired. For several reasons, steady-state nuclear fission detonation waves are not generally appropriate for power production, such as for electrical power generation and the like. Further, nuclear fission deflagration waves are rare in nature, due to having to prevent the initial nuclear fission fuel configuration from disassembling as a hydrodynamic consequence of energy release during the earliest phases of wave propagation. However, in embodiments of the nuclear fission reactor core assembly 100 a nuclear fission deflagration wave can be initiated and propagated in a sub-sonic manner in fissionable fuel whose pressure is substantially independent of its temperature, so that its hydrodynamics is substantially ‘clamped’. The nuclear fission deflagration wave's propagation speed within the nuclear fission reactor core assembly 100 can be controlled in a manner conducive to large-scale civilian power generation, such as in an electricity-producing reactor system like embodiments of the reactor 10. Nucleonics of the nuclear fission deflagration wave are explained below. Inducing nuclear fission of selected isotopes of the actinide elements—the fissile ones—by capture of neutrons of any energy permits the release of nuclear binding energy at any material temperature, including arbitrarily low ones. Release of more than a single neutron per neutron captured, on the average, by nuclear fission of substantially any actinide isotope admits the possibility-in-principle of a diverging neutron-mediated nuclear-fission chain reaction in such materials. Release of more than two neutrons for every neutron which is captured (over certain neutron-energy ranges, on the average) by nuclear fission by some actinide isotopes admits the possibility-in-principle of first converting an atom of a non-fissile isotope to a fissile one (via neutron capture and subsequent beta-decay) by an initial neutron capture, and then of neutron-fissioning the nucleus of the newly-created fissile isotope in the course of a second neutron capture. Most really high-Z (Z≧90) nuclear species can be combusted if, on the average, one neutron from a given nuclear fission event can be radiatively captured on a non-fissile-but-‘fertile’ nucleus which will then convert (such as via beta-decay) into a fissile nucleus and a second neutron from the same fission event can be captured on a fissile nucleus and, thereby, induce fission. In particular, if either of these arrangements is steady-state, then sufficient conditions for propagating a nuclear fission deflagration wave in the given material can be satisfied. Due to beta-decay in the process of converting a fertile nucleus to a fissile nucleus, the characteristic speed of wave advance is of the order of the ratio of the distance traveled by a neutron from its fission-birth to its radiative capture on a fertile nucleus to the half-life of the (longest-lived nucleus in the chain of) beta-decay leading from the fertile nucleus to the fissile one. Since such a characteristic fission neutron-transport distance in normal-density actinides is approximately 10 cm and the beta-decay half-life is 105-106 seconds for most cases of interest, the characteristic wave-speed is 10−4-10−7 cm sec−1, or 10−13-10−14 of that of a nuclear detonation wave. Such a “glacial” speed-of-advance makes clear that the wave is that of a deflagration wave, not of a detonation wave. The deflagration wave propagates not only very slowly but very stably. If such a wave attempts to accelerate, its leading-edge counters ever-more-pure fertile material (which is quite lossy in a neutronic sense), for the concentration of fissile nuclei well ahead of the center of the wave becomes exponentially low, and thus the wave's leading-edge (referred to herein as a “burnfront”) stalls. Conversely, if the wave slows, however, the local concentration of fissile nuclei arising from continuing beta-decay increases, the local rates of fission and neutron production rise, and the wave's leading-edge, that is the burnfront, accelerates. Finally, if the heat associated with nuclear fission is removed sufficiently rapidly from all portions of the configuration of initially fertile matter in which the wave is propagating, the propagation may take place at an arbitrarily low material temperature—although the temperatures of both the neutrons and the fissioning nuclei may be around 1 MeV. Such conditions for initiating and propagating a nuclear fission deflagration wave can be realized with readily available materials. While fissile isotopes of actinide elements are rare terrestrially, both absolutely and relative to fertile isotopes of these elements, fissile isotopes can be concentrated, enriched and synthesized. The use of both naturally-occurring and man-made ones, such as U235 and Pu239, respectively, in initiating and propagating nuclear fission detonation waves is well-known. Consideration of pertinent neutron cross-sections (shown in FIGS. 1B and 1C) suggests that a nuclear fission deflagration wave can burn a large fraction of a core of naturally-occurring actinides, such as Th232 or U238, if the neutron spectrum in the wave is a ‘hard’ or ‘fast’ one. That is, if the neutrons which carry the chain reaction in the wave have energies which are not very small compared to the approximately 1 MeV at which they are evaporated from nascent fission fragments, then relatively large losses to the spacetime-local neutron economy can be avoided when the local mass-fraction of fission products becomes comparable to that of the fertile material (recalling that a single mole of fissile material fission-converts to two moles of fission-product nuclei). Even neutronic losses to typical neutron-reactor structural materials, such as Ta, which has desirable high-temperature properties, may become substantial at neutron energies ≦0.1 MeV. Another consideration is the (comparatively small) variation with incident neutron energy of the neutron multiplicity of fission, ν, and the fraction of all neutron capture events which result in fission (rather than merely γ-ray emission). The algebraic sign of the function α(ν−2) constitutes a necessary condition for the feasibility of nuclear fission deflagration wave propagation in fertile material compared with the overall fissile isotopic mass budget, in the absence of neutron leakage from the core or parasitic absorptions (such as on fission products) within its body, for each of the fissile isotopes of the nuclear fission reactor core assembly 100. The algebraic sign is generally positive for all fissile isotopes of interest, from fission neutron-energies of approximately 1 MeV down into the resonance capture region. The quantity α(ν−2)/ν upper-bounds the fraction of total fission-born neutrons which may be lost to leakage, parasitic absorption, or geometric divergence during deflagration wave propagation. It is noted that this fraction is 0.15-0.30 for the major fissile isotopes over the range of neutron energies which prevails in all effectively unmoderated actinide isotopic configurations of practical interest (approximately 0.1-1.5 MeV). In contrast to the situation prevailing for neutrons of (epi-) thermal energy (see FIG. 1C), in which the parasitic losses due to fission products dominate those of fertile-to-fissile conversion by 1-1.5 decimal orders-of-magnitude, fissile element generation by capture on fertile isotopes is favored over fission-product capture by 0.7-1.5 orders-of-magnitude over the neutron energy range 0.1-1.5 MeV. The former suggests that fertile-to-fissile conversion will be feasible only to the extent of 1.5-5% percent at-or-near thermal neutron energies, while the latter indicates that conversions in excess of 50% may be expected for near-fission energy neutron spectra. In considering conditions for propagation of a nuclear fission deflagration wave, neutron leakage may be effectively ignored for very large, “self-reflected” actinide configurations. Referring to FIG. 1C and analytic estimates of the extent of neutron moderation-by-scattering entirely on actinide nuclei, it will be appreciated that deflagration wave propagation can be established in sufficiently large configurations of the two types of actinides that are relatively abundant terrestrially: Th232 and U238, the exclusive and the principal (that is, longest-lived) isotopic components of naturally-occurring thorium and uranium, respectively. Specifically, transport of fission neutrons in these actinide isotopes will likely result in either capture on a fertile isotopic nucleus or fission of a fissile one before neutron energy has decreased significantly below 0.1 MeV (and thereupon becomes susceptible with non-negligible likelihood to capture on a fission-product nucleus). Referring to FIG. 1B, it will be appreciated that fission product nuclei concentrations must significantly exceed fertile ones and fissile nuclear concentrations may be an order-of-magnitude less than the lesser of fission-product or fertile ones before it becomes quantitatively questionable. Consideration of pertinent neutron scattering cross-sections suggests that right circular cylindrical configurations of actinides which are sufficiently extensive to be effectively infinitely thick—that is, self-reflecting—to fission neutrons in their radial dimension will have density-radius products >>200 gm/cm2—that is, they will have radii >>10-20 cm of solid-density U238—Th232. As an example, studies have indicated that circular cylinders of solid-density Th232 of 25 cm radius, overcoated with an annular shell of 15 cm of C12 (as graphite), may propagate nuclear fission deflagration waves with ≧70% burn-up of the Th232 initially present. Moreover, studies have indicated that replacing the Th232 with half-density U238 may yield similar results—albeit fertile isotope burn-up of ≧80% is realized (as would be expected from inspection of FIG. 1C). A basic condition on the ‘local’ geometry of the breeding-and-burning wave is that the flux history of neutrons excess to the local fissioning process in the core of the burn wave be quantitatively sufficient to at-least-reproduce the fissile atom density 1-2 mean-free-paths into the yet-unburned fuel, in a self-consistent sense. The ‘ash’ behind the burn-wave's peak is substantially ‘neutronically neutral’ in such an accounting scheme, since the neutronic reactivity of its fissile fraction is just balanced by the parasitic absorptions of structure and fission product inventories on top of leakage. If the fissile atom inventory in the wave's center and just in advance of it is time-stationary as the wave propagates, then it's doing so stably; if less, then the wave is ‘dying’, while if more, the wave may be said to be ‘accelerating.’ Thus, a nuclear fission deflagration wave may be propagated and maintained in substantially steady-state conditions for long time intervals in configurations of naturally-occurring actinide isotopes. The above discussion has considered, by way of non-limiting example, circular cylinders of natural uranium or thorium metal of less than a meter or so diameter—and that may be substantially smaller in diameter if efficient neutron reflectors are employed—that may stably propagate nuclear fission deflagration waves for arbitrarily great axial distances. However, propagation of nuclear fission deflagration waves is not to be construed to be limited to circular cylinders, to symmetric geometries, or to singly-connected geometries. To that end, additional embodiments of alternate geometries of the nuclear fission reactor core 100 will be described later. Propagation of a nuclear fission deflagration wave has implications for embodiments of the nuclear fission reactor 10. As a first example, local material temperature feedback can be imposed on the local nuclear reaction rate at an acceptable expense in the deflagration wave's neutron economy. Such a large negative temperature coefficient of neutronic reactivity confers an ability to control the speed-of-advance of the deflagration wave. If very little thermal power is extracted from the burning fuel, its temperature rises and the temperature-dependent reactivity falls, and the nuclear fission rate at wave-center becomes correspondingly small and the wave's equation-of-time reflects only a very small axial rate-of-advance. Similarly, if the thermal power removal rate is large, the material temperature decreases and the neutronic reactivity rises, the intra-wave neutron economy becomes relatively undamped, and the wave advances axially relatively rapidly. Details regarding exemplary implementations of temperature feedback within embodiments of the nuclear fission reactor core assembly 100 will be discussed later. As a second example of implications of propagation of a nuclear fission deflagration wave on embodiments of the nuclear fission reactor 10, less than all of the total fission neutron production in the nuclear fission reactor 10 may be utilized. For example, the local material-temperature thermostating modules may use around 5-10% of the total fission neutron production in the nuclear fission reactor 10. Another ≦10% of the total fission neutron production in the nuclear fission reactor 10 may be lost to parasitic absorption in the relatively large quantities of high-performance, high temperature, structure materials (such as Ta, W, or Re) employed in structural components of the nuclear fission reactor 10. This loss occurs in order to realize ≧60% thermodynamic efficiency in conversion to electricity and to gain high system safety figures-of-merit. The Zs of these materials, such as Ta, W and Re, are approximately 80% of that of the actinides, and thus their radiative capture cross-sections for high-energy neutrons are not particularly small compared to those of the actinides, as is indicated for Ta in FIGS. 1B and 1C. A final 5-10% of the total fission neutron production in the nuclear fission reactor 10 may be lost to parasitic absorption in fission products. As noted above, the neutron economy characteristically is sufficiently rich that approximately 0.7 of total fission neutron production is sufficient to sustain deflagration wave-propagation in the absence of leakage and rapid geometric divergence. This is in sharp contrast with (epi) thermal-neutron power reactors employing low-enrichment fuel, for which neutron-economy discipline in design and operation must be strict. As a third example of implications of propagation of a nuclear fission deflagration wave on embodiments of the nuclear fission reactor 10, high burn-ups (on the order of around 50% to around 80%) of initial actinide fuel-inventories which are characteristic of the nuclear fission deflagration waves permit high-efficiency utilization of as-mined fuel—moreover without a requirement for reprocessing. Referring now to FIGS. 1D-1H, features of the fuel-charge of embodiments of the nuclear fission reactor core assembly 100 are depicted at four equi-spaced times during the operational life of the reactor after origination of the nuclear fission deflagration wave (sometimes referred to herein as “nuclear fission ignition”) in a scenario in which full reactor power is continuously demanded over a ⅓ century time-interval. In the embodiment shown, two nuclear fission deflagration wavefronts propagate from an origination point 28 (near the center of the nuclear fission reactor core assembly 100) toward ends of the nuclear fission reactor core assembly 100. Corresponding positions of the leading edge of the nuclear fission deflagration wave-pair at various time-points after full ignition of the fuel-charge of the nuclear fission reactor core assembly 100 are indicated in FIG. 1D. FIGS. 1E, 1F, 1G, and 1G illustrate masses (in kg of total mass per cm of axial core-length) of various isotopic components in a set of representative near-axial zones and fuel specific power (in W/g) at the indicated axial position as ordinate-values versus axial position along an exemplary, non-limiting 10-meter-length of the fuel-charge as an abscissal value at approximate times after nuclear fission ignition of approximately 7.5 years, 15 years, 22.5 years, and 30 years, respectively. The central perturbation is due to the presence of the nuclear fission igniter module indicated by the origination point 28 (FIG. 1D). It will be noted that the neutron flux from the most intensely burning region behind the burnfront breeds a fissile isotope-rich region at the burnfront's leading-edge, thereby serving to advance the nuclear fission deflagration wave. After the nuclear fission deflagration wave's burnfront has swept over a given mass of fuel, the fissile atom concentration continues to rise for as long as radiative capture of neutrons on available fertile nuclei is considerably more likely than on fission product nuclei, while ongoing fission generates an ever-greater mass of fission products. Nuclear power-production density peaks in this region of the fuel-charge, at any given moment. It will also be noted that in the illustrated embodiments, differing actions of two slightly different types of thermostating units on the left and the right sides of the igniter module account for the corresponding slightly differing power production levels. Still referring to FIGS. 1D-1H, it can be seen that well behind the nuclear fission deflagration wave's advancing burnfront, the concentration ratio of fission product nuclei (whose mass closely averages half that of a fissile nucleus) to fissile ones climbs to a value comparable to the ratio of the fissile fission to the fission product radiative capture cross-sections (FIG. 1B), the “local neutronic reactivity” thereupon goes slightly negative, and both burning and breeding effectively cease—as will be appreciated from comparing FIGS. 1E, 1F, 1G, and 1H with each other, far behind the nuclear fission deflagration wave burnfront. In some embodiments of the nuclear fission reactor 10, all the nuclear fission fuel ever used in the reactor is installed during manufacture of the nuclear fission reactor core assembly 100, and no spent fuel is ever removed from the nuclear fission reactor core assembly 100, which is never accessed after nuclear fission ignition. However, in some other embodiments of the nuclear fission reactor 10, additional nuclear fission fuel is added to the nuclear fission reactor core assembly 100 after nuclear fission ignition. However, in some other embodiments of the nuclear fission reactor 10, spent fuel is removed from the reactor core assembly (and, in some embodiments, removal of spent fuel from the nuclear fission reactor core assembly 100 may be performed while the nuclear fission reactor 10 is operating at power). Regardless of whether or not spent fuel is removed, pre-expansion of the as-loaded fuel permits higher-density actinides to be replaced with lower-density fission products without any overall volume changes in fuel elements, as the nuclear fission deflagration wave sweeps over any given axial element of actinide ‘fuel,’ converting it into fission-product ‘ash.’ Launching of nuclear fission deflagration waves into Th232 or U238 fuel-charges is readily accomplished with ‘nuclear fission igniter modules’ enriched in fissile isotopes. Higher enrichments result in more compact modules, and minimum mass modules may employ moderator concentration gradients. In addition, nuclear fission igniter module design may be determined in part by non-technical considerations, such as resistance to materials diversion for military purposes in various scenarios. Such modules may employ U235 in U238, in sufficiently low concentration as to be effectively non-detonatable in any quantity or configuration—such as ≦20%—in contrast, for example, to technically more optimal Pu239 in Th232. Quantities of U235 already excess to military stockpiles suffice for ≧104 such nuclear fission igniter modules, corresponding to a total inventory of nuclear fission power reactors sufficient to supply 10 billion people with kilowatt-per-capita electricity. While the illustrative nuclear fission igniter of the previously described embodiments included nuclear fission material configured to initiate propagation of the burning wavefront, in other approaches, the nuclear fission igniter may include other types of reactivity sources in addition to or in place of those previously described. For example, nuclear fission igniters may include “burning embers”, e.g., nuclear fission fuel enriched in fissile isotopes via exposure to neutrons within a propagating nuclear fission deflagration wave reactor. Such “burning embers” may function as nuclear fission igniters, despite the presence of various amounts of fission products “ash”. For example, nuclear fission igniters may include neutron sources using electrically driven sources of high energy ions (such as protons, deuterons, alpha particles, or the like) or electrons that may in turn produce neutrons. In one illustrative approach, a particle accelerator, such as a linear accelerator may be positioned to provide high energy protons to an intermediate material that may in turn provide such neutrons (e.g., through spallation). In another illustrative approach, a particle accelerator, such as a linear accelerator may be positioned to provide high energy electrons to an intermediate material that may in turn provide such neutrons (e.g., by electro-fission and/or photofission of high-Z elements). Alternatively, other known neutron emissive processes and structures, such as electrically induced fusion approaches, may provide neutrons (e.g., 14 Mev neutrons from D-T fusion) that may thereby initiate the propagating fission wave. Now that nucleonics of the fuel charge and the nuclear fission deflagration wave have been discussed, further details regarding “nuclear fission ignition” and maintenance of the nuclear fission deflagration wave will be discussed. A centrally-positioned nuclear fission igniter moderately enriched in fissionable material, such as U235, has a neutron-absorbing material (such as a borohydride) removed from it (such as by operator-commanded electrical heating), and the nuclear fission igniter becomes neutronically critical. Local fuel temperature rises to a design set-point and is regulated thereafter by the local thermostating modules (discussed in detail later). Neutrons from the fast fission of U235 are mostly captured at first on local U238 or Th232. It will be appreciated that uranium enrichment of the nuclear fission igniter may be reduced to levels not much greater than that of light water reactor (LWR) fuel by introduction into the nuclear fission igniter and the fuel region immediately surrounding it of a radial density gradient of a refractory moderator, such as graphite. High moderator density enables low-enrichment fuel to burn satisfactorily, while decreasing moderator density permits efficient fissile breeding to occur. Thus, optimum nuclear fission igniter design may involve trade-offs between proliferation robustness and the minimum latency from initial criticality to the availability of full-rated-power from the fully-ignited fuel-charge of the core. Lower nuclear fission igniter enrichments entail more breeding generations and thus impose longer latencies. The maximum (unregulated) reactivity of the nuclear fission reactor core assembly 100 slowly decreases in the first phase of the nuclear fission ignition process because, although the total fissile isotope inventory is increasing monotonically, this total inventory is becoming more spatially dispersed. As a result of choice of initial fuel geometry, fuel enrichment versus position, and fuel density, it may be arranged for the maximum reactivity to still be slightly positive at the time-point at which its minimum value is attained. Soon thereafter, the maximum reactivity begins to increase rapidly toward its greatest value, corresponding to the fissile isotope inventory in the region of breeding substantially exceeding that remaining in the nuclear fission igniter. A quasi-spherical annular shell then provides maximum specific power production. At this point, the fuel-charge of the nuclear fission reactor core assembly 100 is referred to as “ignited.” Now that the fuel-charge of the nuclear fission reactor core assembly 100 has been “ignited”, propagation of the nuclear fission deflagration wave, also referred to herein as “nuclear fission burning”, will now be discussed. The spherically-diverging shell of maximum specific nuclear power production continues to advance radially from the nuclear fission igniter toward the outer surface of the fuel charge. When it reaches this surface, it naturally breaks into two spherical zonal surfaces, with one surface propagating in each of the two opposite directions along the axis of the cylinder. At this time-point, the full thermal power production potential of the core has been developed. This epoch is characterized as that of the launching of the two axially-propagating nuclear fission deflagration wave burnfronts. In some embodiments the center of the core's fuel-charge is ignited, thus generating two oppositely-propagating waves. This arrangement doubles the mass and volume of the core in which power production occurs at any given time, and thus decreases by two-fold the core's peak specific power generation, thereby quantitatively minimizing thermal transport challenges. However, in other embodiments, the core's fuel charge is ignited at one end, as desired for a particular application. In other embodiments, the core's fuel charge may be ignited in multiple sites. In yet other embodiments, the core's fuel charge is ignited at any 3-D location within the core as desired for a particular application. In some embodiments, two propagating nuclear fission deflagration waves will be initiated and propagate away from a nuclear fission ignition site, however, depending upon geometry, nuclear fission fuel composition, the action of neutron modifying control structures or other considerations, different numbers (e.g., one, three, or more) of nuclear fission deflagration waves may be initiated and propagated. However, for sake of understanding, the discussion herein refers, without limitation, to propagation of two nuclear fission deflagration wave burnfronts. From this time forward through the break-out of the two waves when they reach the two opposite ends, the physics of nuclear power generation is effectively time-stationary in the frame of either wave, as illustrated in FIGS. 1E-1H. The speed of wave advance through the fuel is proportional to the local neutron flux, which in turn is linearly dependent on the thermal power demanded from the nuclear fission reactor core assembly 100 via the collective action on the nuclear fission deflagration wave's neutron budget of the thermostating modules (not shown). When more power is demanded from the reactor via lower-temperature coolant flowing into the core, the temperature of the two ends of the core (which in some embodiments are closest to the coolant inlets) decreases slightly below the thermostating modules' design set-point, a neutron absorber is thereby withdrawn from the corresponding sub-population of the core's thermostating modules, and the local neutron flux is permitted thereby to increase to bring the local thermal power production to the level which drives the local material temperature up to the set-point of the local thermostating modules. However, in the two burnfront embodiment this process is not effective in heating the coolant significantly until its two divided flows move into the two nuclear burn-fronts. These two portions of the core's fuel-charge—which are capable of producing significant levels of nuclear power when not suppressed by the neutron absorbers of the thermostating modules—then act to heat the coolant to the temperature specified by the design set-point of their modules, provided that the nuclear fission fuel temperature does not become excessive (and regardless of the temperature at which the coolant arrived in the core). The two coolant flows then move through the two sections of already-burned fuel centerward of the two burnfronts, removing residual nuclear fission and afterheat thermal power from them, both exiting the fuel-charge at its center. This arrangement encourages the propagation of the two burnfronts toward the two ends of the fuel-charge by “trimming” excess neutrons primarily from the trailing edge of each front, as illustrated in FIGS. 1E-1H. Thus, the core's neutronics may be considered to be substantially self-regulated. For example, for cylindrical core embodiments, the core's nucleonics may be considered to be substantially self-regulating when the fuel density-radius product of the cylindrical core is ≧200 gm/cm2 (that is, 1-2 mean free paths for neutron-induced fission in a core of typical composition, for a reasonably fast neutron spectrum). The primary function of the neutron reflector in such core designs is to drastically reduce the fast neutron fluence seen by the outer portions of the reactor, such as its radiation shield, structural supports, thermostating modules and outermost shell. Its incidental influence on the performance of the core is to improve the breeding efficiency and the specific power in the outermost portions of the fuel, though the value of this is primarily an enhancement of the reactor's economic efficiency. Outlying portions of the fuel-charge are not used at low overall energetic efficiency, but have isotopic burn-up levels comparable to those at the center of the fuel-charge. Final, irreversible negation of the core's neutronic reactivity may be performed at any time by injection of neutronic poison into the coolant stream, via either the primary loops which extend to the application heat exchangers 16 (FIG. 1A) or the afterheat-dumping loops connecting the nuclear fission reactor 10 (FIG. 1A) to the heat dump heat exchangers 26 (FIG. 1A). For example, lightly loading the coolant stream with a material such as BF3, possibly accompanied by a volatile reducing agent such as H2 if desired, may deposit metallic boron substantially uniformly over the inner walls of the coolant-tubes threading through the reactor's core, via exponential acceleration of the otherwise slow chemical reaction 2BF3+3H2→2B+6HF by the high temperatures found therein. Boron, in turn, is a highly refractory metalloid, and will not migrate from its site of deposition. Substantially uniform presence of boron in the core in <100 kg quantities may negate the core's neutronic reactivity for indefinitely prolonged intervals without involving the use of powered mechanisms in the vicinity of the reactor. Exemplary Embodiments and Aspects of Reactor Core Assemblies Exemplary embodiments and aspects of the nuclear fission reactor core assembly 100 and exemplary nuclear fission fuel charges disposed therein will now be discussed. Referring now to FIG. 1I, the nuclear fission reactor core assembly 100 is suitable for use with a fast neutron spectrum nuclear fission reactor. It will be appreciated that the nuclear fission reactor core assembly 100 is shown schematically in FIG. 1I. As such, no geometric limitations are intended regarding shape of the nuclear fission reactor core assembly 100. As mentioned above, details were discussed for circular cylinders of natural uranium or thorium metal that may stably propagate nuclear fission deflagration waves for arbitrarily great axial distances. However, it is again emphasized that propagation of nuclear fission deflagration waves is not to be construed to be limited to circular cylinders or to metallic nuclear fission fuels, or to pure uranium or thorium nuclear fission fuel materials. To that end, additional embodiments of alternate geometries of the nuclear fission reactor core assembly 100 and fuel charges disposed therein will be described later. A neutron reflector/radiation shield 120 surrounds nuclear fission fuel 130. The nuclear fission fuel 130 is fissionable material, that is material appropriate for undergoing fission in a nuclear fission reactor, examples of which are actinide or transuranic elements. As discussed above, the fissionable material for the nuclear fission fuel 130 may include without limitation Th232 or U238. However, in other embodiments discussed below, other fissionable material may be used in the nuclear fission fuel 130. In some embodiments, the nuclear fission fuel 130 is contiguous. In other embodiments, the nuclear fission fuel 130 is non-contiguous. A nuclear fission igniter 110 acts within the nuclear fission fuel 130 for initiating a nuclear fission deflagration wave burnfront (not shown). The nuclear fission igniter 110 is made and operates according to principles and details discussed above. Therefore, details of construction and operation of the nuclear fission igniter 110 need not be repeated for sake of brevity. Referring now to FIG. 1J, after the nuclear fission fuel 130 (FIG. 1I) has been ignited by the nuclear fission igniter 110 (in a manner as discussed above), a propagating burnfront 140 (that is, a propagating nuclear fission deflagration wave burnfront, as discussed above) is initiated and propagates throughout the nuclear fission fuel 130 (FIG. 1I) a direction shown by an arrow 144. As discussed above, a region 150 of maximum reactivity is established around the propagating burnfront 140. The propagating burnfront 140 propagates through unburnt nuclear fission fuel 154 in the direction indicated by the arrow 144, leaving behind the propagating burnfront 140 burnt nuclear fission fuel 160 that includes fission products 164, such as isotopes of iodine, cesium, strontium, xenon, and/or barium (and referred to in the discussion above as “fission product ash”). In the context of burnt nuclear fission fuel and unburnt nuclear fission fuel, the term “burning” (as applied to nuclear fission fuel) means that at least some components of the nuclear fission fuel undergo neutron-mediated nuclear fission. In the context of propagating nuclear fission deflagration wave burnfronts, the terms “burning” and “burnt” also mean that at least some components of the nuclear fission fuel undergo “breeding”, whereby neutron absorption is followed by multi-second half-life beta-decay transmutation into one or more fissile isotopes, which then may or may not undergo neutron-mediated nuclear fission. Thus, the unburnt nuclear fission fuel 154 may be considered a first neutron environment having a first set of neutron environment parameters. Similarly, the burnt nuclear fission fuel 160 may be considered a second neutron environment having a second set of neutron environment parameters that are different than the first set of neutron environment parameters. The term “neutron environment” refers to the detailed neutron distribution, including its variation with respect to time, space, direction, and energy. The neutron environment includes the aggregate of multiple individual neutrons, each of which may occupy different locations at different times, and each of which may have different directions of motion and different energies. In some circumstances, a nuclear environment may be characterized by a reduced subset of these detailed properties. In one example, a reduced subset may include an aggregation of all neutrons within given space, time, direction, and energy ranges of specified time, space, direction, and energy values. In another example, some or all of the time, space, direction, or energy aggregations may incorporate value-dependent weighting functions. In another example, a reduced subset may include weighted aggregation over the full range of direction and energy values. In another example, the aggregation over energies may involve energy-dependent weighting by a specified energy function. Examples of such weighting functions include material and energy-dependent cross-sections, such as those for neutron absorption or fission. In some embodiments, only the propagating burnfront 140 is originated and propagated through the unburnt nuclear fission fuel 154. In such embodiments, the nuclear fission igniter 110 may be located as desired. For example, the nuclear fission igniter 110 may be located toward the center of the nuclear fission fuel 130 (FIG. 1I). In other embodiments (not shown) the nuclear fission igniter 110 may be located toward an end of the nuclear fission fuel 130. In other embodiments, in addition to the propagating burnfront 140, a propagating burnfront 141 is originated and propagated through the other fuel 154 along a direction indicated by an arrow 145. A region 151 of maximum reactivity is established around the obligating burnfront 141. The propagating burnfront 141 leaves behind it the burnt nuclear fission fuel 160 and the fission products 164. Principles and details of origination and propagation of the propagating burnfront 141 are the same as that previously discussed for the propagating burnfront 140. Therefore, details of origination and propagation of the propagating burnfront 141 need not be provided for sake of brevity. Referring now to FIG. 2A, a nuclear fission reactor 200, such as a fast neutron spectrum nuclear fission reactor, includes nuclear fission fuel assemblies 210 disposed therein. The following discussion includes details of exemplary nuclear fission fuel assemblies 210 that may be used in the nuclear fission reactor 200. Other details regarding the nuclear fission reactor 200, including origination and propagation of a nuclear fission deflagration wave burnfront (that is, “burning” the nuclear fission fuel) are similar to those of the nuclear fission reactor 10 (FIG. 1A), and need not be repeated for sake of brevity. Referring now to FIG. 2B and given by way of non-limiting example, in one embodiment the nuclear fission fuel assembly 210 suitably includes a previously burnt nuclear fission fuel assembly 220. The previously burnt nuclear fission fuel assembly 220 is clad with cladding 224. The cladding 224 is the “original” cladding in which the previously burned nuclear fission fuel assembly 220 was clad. The term “previously burnt” means that at least some components of the nuclear fission fuel assembly have undergone neutron-mediated nuclear fission and that the isotopic composition of the nuclear fission fuel has been modified. That is, the nuclear fission fuel assembly has been put in a neutron spectrum or flux (either fast or slow), at least some components have undergone neutron-mediated nuclear fission and, as result, the isotopic composition of the nuclear fission fuel has been changed. Thus, a burnt nuclear fission fuel assembly 220 may have been previously burnt in any reactor, such as without limitation a light water reactor. It is intended that the previously burnt nuclear fission fuel assembly 220 can include without limitation any type of nuclear fissionable material whatsoever appropriate for undergoing fission in a nuclear fission reactor, such as actinide or transuranic elements like natural thorium, natural uranium, enriched uranium, or the like. In some other embodiments, the previously burnt nuclear fission fuel assembly 220 may not be clad with “original” cladding 224, but in these embodiments, the previously burnt nuclear fission fuel assembly 220 is chemically untreated subsequent to its previous burning in the nuclear fission reactor 200. Referring now to FIG. 2C, the previously burnt nuclear fission fuel assembly 220 and its “original” cladding 224 is clad with cladding 230. Thus, the previously burnt nuclear fission fuel assembly 220 is retained in its original cladding 224, and the cladding 230 is disposed around an exterior of the cladding 224. The cladding 230 can accommodate swelling. For example, when the previously burnt nuclear fission fuel assembly 220 was burnt in a light water reactor, the cladding 224 was sufficient to contain swelling at approximately 3% burn-up of the previously burnt nuclear fission fuel assembly 220. In one nonlimiting example, the cladding 230 contacts the cladding 224 at azimuthally, symmetric, cylindrical faces around the cladding 224. Such an arrangement enables removal of heat through the contacting faces while allowing at least one half of the cladding 224 to expand into void spaces between the cladding 224 and the cladding 230. In some embodiments, the cladding 230 is made up of cladding sections (not shown) that are configured to help accommodate swelling into the void spaces, as described above. In other embodiments, the cladding 230 may be provided as a barrier, such as a tube, provided between an exterior of the cladding 224 and reactor coolant (not shown). In some other embodiments, the previously burned nuclear fission fuel assembly 220 is burnt in the nuclear fission reactor 200 as the nuclear fission fuel assembly 210. That is, the previously burnt nuclear fission fuel assembly 220 may not be clad with the cladding 230. This embodiment envisions burning the previously burnt nuclear fission fuel assembly 220, such as one that was burnt in a light water reactor, or in a fast neutron spectrum nuclear fission reactor, or in any other form of nuclear fission reactor and either (a) tolerating or planning to accept possible failure of the cladding 224 due to swell or, (b) burning the previously burnt nuclear fission fuel assembly 220 in the fast neutron spectrum nuclear fission reactor 200 to levels significantly less than isotopic depletion (in which case swelling may be of acceptable magnitude). Referring now to FIGS. 3A, 3B, 3C, and 3D, alternate nuclear fission fuel geometries of nuclear fission fuel structures 310, 320, 330, and 340, respectively, are discussed. Each of the nuclear fission fuel structures 310, 320, 330, and 340 includes a nuclear fission igniter 300, and a propagating nuclear fission deflagration wave 302 is propagated in a direction indicated by an arrow 304. In a spherical nuclear fission fuel structure 310 (FIG. 3A), the nuclear fission igniter 300 is disposed toward a center of the spherical nuclear fission fuel structure 310. The propagating burnfront 302 propagates radially outward from the nuclear fission igniter 300, as indicated by the arrows 304. In a parallelepiped nuclear fission fuel structure 320, the nuclear fission igniter 300 is disposed as desired. As discussed above, two propagating burnfronts 302 may be originated and propagated toward ends of the parallelepiped nuclear fission fuel structure 320 along directions indicated by the arrows 304. Alternately, the nuclear fission igniter 300 may be disposed toward an end of the parallelepiped nuclear fission fuel structure 320, in which case one propagating burnfront 302 is originated and propagates toward the other end of the parallelepiped nuclear fission fuel structure 320 along the direction indicated by the arrow 304. In a toroidal nuclear fission fuel structure 330 (FIG. 3C), the nuclear fission igniter 300 is disposed as desired. Two propagating burnfronts 302 may be originated and propagated away from the nuclear fission igniter 300 and toward each other along directions indicated by the arrows 304. In such a case, the toroidal nuclear fission fuel structure 330 may be considered to be “burnt” when the propagating burnfronts 302 meet, and propagation of the propagating burnfront 302 may stop. Alternately, only one propagating burnfront 302 is originated and propagates around the toroidal nuclear fission fuel structure 330 along the direction indicated by the arrow 304. In such a case, the toroidal nuclear fission fuel structure 330 may be considered to be “burnt” when the propagating burnfront 302 returns to the site of the nuclear fission igniter 300, and propagation of the propagating burnfront 302 may stop or may be re-started. In another embodiment, the propagating burnfront 302 is “restarted” due to the removal or decay of fission products during the burnfront's propagation around the toroid. In another embodiment, the propagating burnfront 302 is “restarted” due to control of neutron modifying structures, as discussed later. In another embodiment, the toroidal nuclear fission fuel structure 330 is not a “geometric” toroid, but a “logical” toroid, with a more general reentrant structure. As mentioned above, nuclear fission deflagration propagating wave burnfronts can be initiated and propagated in nuclear fission fuels having any shape as desired. For example, in an irregularly-shaped nuclear fission fuel structure 340, the nuclear fission igniter 300 can be located as desired. Propagating burnfronts 302 are initiated and propagate along directions indicated by the arrows 304 as desired for a particular application. In one approach, thermal management may be adjusted to provide thermal control appropriate for any changes in operational parameters, such as revised neutronic action of the previously burnt or modified nuclear fission fuel or other parameter changes, that may result from removal of ash, addition of fuels, or from other parameters of re-burning. In these exemplary geometries, the nuclear fission ignitor 300 may be any of the varieties of nuclear fission ignitor previously discussed. The indicated nuclear fission ignitor 300 is the site at which nuclear fission ignition occurs, but for some embodiments (e.g., electrical neutron sources) additional components of the nuclear fission ignitor may exist, and may reside in different physical locations. Referring now to FIG. 4, a nuclear fission fuel structure 400 includes a nuclear fission igniter 410 and non-contiguous segments 420 of nuclear fission fuel material. The behavior of a nuclear fission deflagration wave with non-contiguous segments 420 of nuclear fission fuel material is similar to that previously discussed for contiguous nuclear fission fuel material; it is crucial only that the non-continguous segments 420 be in “neutronic” contact, not physical contact. Referring now to FIG. 5, a modular nuclear fission fuel core 500 includes a neutron reflector/radiation shield 510 and modular nuclear fission fuel assemblies 520. The modular nuclear fission fuel assemblies 520 are placed as desired within the fuel assembly receptacles 530. The modular nuclear fission fuel core 500 may be operated in any number of ways. For example, all of the fuel assembly receptacles 530 in the modular nuclear fission fuel core 500 may be fully populated with modular nuclear fission fuel assemblies 520 prior to initial operation (e.g., prior to initial origination and propagation of a nuclear fission deflagration propagating wave burnfont within and through the modular nuclear fission fuel assemblies 520). As another example, after a nuclear fission deflagration wave burnfront has completely propagated through modular nuclear fission fuel assemblies 520, such “burnt” modular nuclear fission fuel assemblies 520 may be removed from their respective fuel assembly receptacles 530 and replaced with unused modular nuclear fission fuel assemblies 540, as desired; this emplacement is indicated by the arrow 544. A nuclear fission deflagration wave burnfront can be initiated in the unused modular nuclear fission fuel assemblies 540, thereby enabling continued or extended operation of the modular nuclear fission fuel core 500 as desired. As another example, the modular nuclear fission fuel core 500 need not be fully populated with modular nuclear fission fuel assemblies 520 prior to initial operation. For example, less than all of the fuel assembly receptacles 530 can be populated with modular nuclear fission fuel assemblies 520. In such a case, the number of modular nuclear fission fuel assemblies 520 that are placed within the modular nuclear fission fuel core 500 can be determined based upon power demand, such as electrical loading in watts, that will be placed upon the modular nuclear fission fuel core 500. A nuclear fission deflagration wave burnfront is originated and propagated through the modular nuclear fission fuel assemblies 520 as previously described. In one approach, thermal management may be adjusted to provide thermal control appropriate to maintain the inserted fuel assembly receptacles 530 at appropriate temperatures. As another example, the modular nuclear fission fuel core 500 again need not be fully populated with modular nuclear fission fuel assemblies 520 prior to initial operation. The number of modular nuclear fission fuel assemblies 520 provided may be determined based upon a number of modular nuclear fission fuel assemblies 520 that are available or for other reasons. A nuclear fission deflagration wave burnfront is originated and propagates through the modular nuclear fission fuel assemblies 520. As the nuclear fission deflagration wave burnfront approaches unpopulated fuel assembly receptacles 530, the unpopulated fuel assembly receptacles 530 can be populated with modular nuclear fission fuel assemblies 520, such as on a “just-in-time” basis; this emplacement is indicated by the arrow 544. Thus, continued or extended operation of the modular nuclear fission fuel core 500 can be enabled without initially fueling the entire modular nuclear fission fuel core 500 with modular nuclear fission fuel assemblies 520. It will be appreciated that the concept of modularity can be extended. For example, in other embodiments, a modular nuclear fission reactor can be populated with any number of nuclear fission reactor cores in the same manner that the modular nuclear fission fuel core 500 can be populated with any number of modular nuclear fission fuel assemblies 520. To that end, the modular nuclear fission reactor can be analogized to the modular nuclear fission fuel core 500 and nuclear fission reactor cores can be analogized to the modular nuclear fission fuel assemblies 520. The several contemplated modes of operation discussed above for the modular nuclear fission fuel core 500 thus apply by analogy to a modular nuclear fission reactor. Applications of modular designs are shown in FIGS. 6A-6C. Referring to FIG. 6A, a nuclear fission facility 600 includes a fast neutron spectrum nuclear fission core assembly 610 that is operationally coupled to an operational sub system 620 (such as without limitation an electrical power generating facility) via a core-subsystem coupling 630 (such as without limitation a reactor coolant system such as a primary loop and, if desired, a secondary loop including a steam generator). Referring now to FIG. 6B, another fast neutron spectrum nuclear fission core assembly 610 may be emplaced within the nuclear fission facility 600. The additional fast neutron spectrum nuclear fission core assembly 610 is operationally coupled to another operational sub system 620 by another core-subsystem coupling 630. The operational sub system's 620 are coupled to each other via a subsystem-subsystem coupling 640. A subsystem-subsystem coupling 640 can provide prime mover or other energy transfer medium between the operational sub systems 620. To that end, energy produced by any one of the nuclear core assemblies 610 can be transferred to any operational sub system 620 as desired. Referring now to FIG. 6C, a third fast neutron spectrum nuclear fission core assembly 610, and associated operational sub system 620, and core-subsystem coupling 630 have been placed in the nuclear fission facility 600. Again, as described above, energy produced by any one of the fast neutron spectrum nuclear fission core assemblies 610 can be transferred to any operational sub system 620 as desired. In other embodiments, this linking process can be more general than discussed above, so that, the nuclear fission facility 600 may consist of a number N of fast neutron spectrum nuclear fission core assemblies 610, and a same or different number M of operational subsystems 620. It will be appreciated, that the individual nuclear fast neutron spectrum nuclear fission core assemblies 610 need not be identical to each other, nor need the operational sub systems 620 be identical to each other. Similarly, the core-subsystem couplings 630 need not be identical to each other, nor do the subsystem-subsystem couplings 640 need be identical to each other. In addition to the operational sub system 620 embodiment discussed above, other embodiments of operational sub system 620 include, without limitation, reactor coolant systems, electrical nuclear fission ignitors, afterlife heat-dumps, reactor site facilities (such as basing and security), and the like. Referring now to FIG. 7, heat energy can be extracted from a nuclear fission reactor core according to another embodiment. In a nuclear fission reactor 700, a nuclear fission deflagration wave burnfront is initiated and propagated in a burning wavefront heat generating region 720, in a manner as described above. Heat absorbing material 710, such as a condensed phase density fluid (e.g., water, liquid metals, terphenyls, polyphenyls, fluorocarbons, FLIBE (2LiF—BeF2) and the like) flows through the region 720 as indicated by an arrow 750, and heat is transferred from the propagating burnfront fission to the heat absorbing material 710. In some fast fission spectrum nuclear reactors, the heat absorbing material 710 is chosen to be a nuclear inert material (such as He4) so as to minimally perturb the neutron spectrum. In some embodiments of the nuclear fission reactor 700, the neutron content is sufficiently robust, so that a non-nuclear-inert heat absorbing material 710 may be acceptably utilized. The heat absorbing material 710 flows to a heat extraction region 730 that is substantially out of thermal contact with the burning wavefront heat generating region 720. The energy 740 is extracted from the heat absorbing material 710 at the heat extraction region 730. The heat absorbing material 710 can reside in either a liquid state, a multiphase state, or a substantially gaseous state upon extraction of the heat energy 740 in the heat extraction region 730. Referring now to FIG. 8, in some embodiments a nuclear fission deflagration wave burnfront can be driven into areas of nuclear fission fuel as desired, thereby enabling a variable nuclear fission fuel burn-up. In a propagating burnfront nuclear fission reactor 800, a nuclear fission deflagration wave burnfront 810 is initiated and propagated as described above. Actively controllable neutron modifying structures 830 can direct or move the burnfront 810 in directions indicated by areas 820. In one embodiment, the actively controllable neutron modifying structures 830 insert neutron absorbers, such as without limitation Li6, B10, or Gd, into nuclear fission fuel behind the burnfront 810, thereby driving down or lowering neutronic reactivity of fuel that is presently being burned by the burnfront 810 relative to neutronic reactivity of fuel ahead of the burnfront 810, thereby speeding up the propagation rate of the nuclear fission deflagration wave. In another embodiment, the actively controllable neutron modifying structures 830 insert neutron absorbers into nuclear fission fuel ahead of the burnfront 810, thereby slowing down the propagation of the nuclear fission deflagration wave. In other embodiments the actively controllable neutron modifying structures 830 insert neutron absorbers into nuclear fission fuel within or to the side of the burnfront 810, thereby changing the effective size of the burnfront 810. In another embodiment, the actively controllable neutron modifying structures 830 insert neutron moderators, such as without limitation hydrocarbons or Li7, thereby modifying the neutron energy spectrum, and thereby changing the neutronic reactivity of nuclear fission fuel that is presently being burned by the burnfront 810 relative to neutronic reactivity of nuclear fission fuel ahead of or behind the burnfront 810. In some situations, an effect of the neutron moderators is associated with detailed changes in the neutron energy spectrum (e.g., hitting or missing cross-section resonances), while in other cases the effects are associated with lowering the mean neutron energy of the neutron environment (e.g., downshifting from “fast” neutron energies to epithermal or thermal neutron energies). In yet other situations, an effect of the neutron moderators is to deflect neutrons to or away from selected locations. In some embodiments, one of the aforementioned effects of neutron moderators is of primary importance, while in other embodiments, multiple effects are of comparable design significance. In another embodiment, the actively controllable neutron modifying structures 830 contain both neutron absorbers and neutron moderators; in one nonlimiting example, the location of neutron absorbing material relative to that of neutron moderating material is changed to affect control (e.g., by masking or unmasking absorbers, or by spectral-shifting to increase or decrease the absorption of absorbers), in another nonlimiting example, control is affected by changing the amounts of neutron absorbing material and/or neutron moderating material. The burnfront 810 can be directed as desired according to selected propagation parameters. For example, propagation parameters can include a propagation direction or orientation of the burnfront 810, a propagation rate of the burnfont 810, power demand parameters such the heat generation density, cross-sectional dimensions of a burning region through which the burnfront 810 is to the propagated (such as an axial or lateral dimension of the burning region relative to an axis of propagation of the burnfront 810), or the like. For example, the propagation parameters may be selected so as to control the spatial or temporal location of the burnfront 810, so as to avoid failed or malfunctioning control elements (e.g., neutron modifying structures or thermostats), or the like. Referring now to FIGS. 9A and 9B, a nuclear fission reactor can be controlled with programmable thermostats, thereby enabling the temperature of the reactor's fuel-charge to be varied over time responsive to changes in operating parameters. Temperature profiles 940 are determined as a function of position through a fuel-charge of a nuclear fission reactor 900. An operating temperature profile 942 of operating temperatures throughout the nuclear fission reactor 900 is established responsive to a first set of operating parameters, such as predicted power draw, thermal creep of structural materials, etc. At other times, or in other circumstances, the operating parameters may be revised. To that end, a revised operating temperature profile 944 of revised operating temperatures throughout the nuclear fission reactor 900 is established. The nuclear fission reactor 900 includes programmable temperature responsive neutron modifying structures 930. The programmable temperature responsive neutron modifying structures 930 (an example of which is described in detail later) introduce and remove neutron absorbing or neutron moderating material into and from the fuel-charge of a nuclear fission reactor 900. A nuclear fission deflagration wave burnfront 910 is initiated and propagated in a fuel-charge of the nuclear fission reactor 900. Responsive to the revised operating temperature profile 944, the programmable temperature responsive neutron modifying structures 930 introduce neutron absorbing or moderating material into the fuel-charge of the nuclear fission reactor 900 to lower operating temperature in the nuclear fission reactor 900 or remove neutron absorbing or moderating material from the fuel-charge of the nuclear fission reactor 900 in order to raise operating temperature of the nuclear fission reactor 900. It will be appreciated, that operating temperature profiles are only one example of control parameters which can be used to determine the control settings of programmable temperature responsive neutron modifying structures 930, which are in such cases responsive to the selected control parameters, not necessarily to the temperature. Nonlimiting examples of other control parameters which can be used to determine the control settings of programmable temperature responsive neutron modifying structures 930 include power levels, neutron levels, neutron spectrum, neutron absorption, fuel burnup levels, and the like. In one example, the neutron modifying structures 930 are used to control fuel burnup levels to relatively low (e.g., <50%) levels in order to achieve high-rate “breeding” of nuclear fission fuel for use in other nuclear fission reactors, or to enhance suitability of the burnt nuclear fission fuel for subsequent re-propagation of a nuclear fission deflagration wave in a propagating nuclear fission deflagration wave reactor. Different control parameters can be used at different times, or in different portions of the reactor. It will be appreciated that the various neutron modifying methods discussed previously in the context of neutron modifying structures can also be utilized in programmable temperature responsive neutron modifying structures 930, including without limitation, the use of neutron absorbers, neutron moderators, combinations of neutron absorbers and/or neutron moderators, variable geometry neutron modifiers, and the like. According to other embodiments and referring now to FIGS. 10A and 10B, material can be nuclearly processed. As shown in FIG. 10A, nuclearly processable material 1020 (that has a set of non-irradiated properties) is placed in a propagating nuclear fission deflagration wave reactor 1000. A nuclear fission deflagration wave propagating burnfront 1030 is originated and propagated along a direction indicated by arrows 1040 as described above. The material 1020 is placed in neutronic coupling with a region of maximized reactivity 1010, that is the material is neutron irradiated, as the nuclear fission deflagration wave propagating burnfront 1030 propagates through or in the vicinity of the material 1020, thereby irradiating the material 1020 and conferring upon the material 1020 a desired set of modified properties. In one embodiment, the neutron irradiation of material 1020 may be controlled by the duration and/or extent of the nuclear fission deflagration wave propagating burnfront 1030. In another embodiment, the neutron irradiation of material 1020 may be controlled by control of the neutron environment (e.g., the neutron energy spectrum for Np237 processing) via neutron modifying structures. In another embodiment, the propagating nuclear fission deflagration wave reactor 1000 may be operated in a “safe” sub-critical manner, relying upon an external source of neutrons to sustain the propagating burnfront 1030, while using a portion of the fission-generated neutrons for nuclear processing of the material 1020. In some embodiments, the material 1020 may be present before nuclear fission ignition occurs within the propagating nuclear fission deflagration wave reactor 1000, while in other embodiments the material 1020 may be added after nuclear fission ignition. In some embodiments, the material 1020 is removed from the propagating nuclear fission deflagration wave reactor 1000, while in other embodiments it remains in place. Alternately and as shown in FIG. 10B, a nuclear fission deflagration wave propagating burnfront 1030 is initiated and propagated in a propagating nuclear fission deflagration wave reactor 1000 along a direction indicated by arrows 1040. Material 1050 having a set of non-irradiated properties is loaded into the propagating nuclear fission deflagration wave reactor 1000. As indicated generally at 1052, the material 1050 in transported into physical proximity and neutronic coupling with a region of maximized reactivity as the nuclear fission deflagration wave propagating burnfront 1030 passes through the material 1050. The material 1050 remains in neutronic coupling for a sufficient time interval to convert the material 1050 into material 1056 having a desired set of modified properties. Upon the material 1050 having thus been converted into the material 1056, the material 1056 may be physically transported out of the reactor 1000 as generally indicated at 1054. The removal 1054 can take place either during operation of the propagating nuclear fission deflagration wave reactor 1000 or afterward it has been “shut-off”, and can be performed in either a continuous, sequential, or batch process. In one example, the nuclearly processed material 1056 may be subsequently used as nuclear fission fuel in another nuclear fission reactor, such as without limitation LWRs or propagating nuclear fission deflagration wave reactors. In another nonlimiting example, the nuclearly processed material 1056 may be subsequently used within the nuclear fission ignitor of a propagating nuclear fission deflagration wave reactor. In one approach, thermal management may be adjusted to provide thermal control appropriate for any changes in operational parameters, as appropriate for the revised materials or structures. According to further embodiments, temperature-driven neutron absorption can be used to control a nuclear fission reactor, thereby “engineering-in” an inherently-stable negative temperature coefficient of reactivity (αT). Referring now to FIG. 11A, a nuclear fission reactor 1100 is instrumented with temperature detectors 1110, such as without limitation thermocouples. In this embodiment the nuclear fission reactor 1100 suitably can be any type of fission reactor whatsoever. To that end, the nuclear fission reactor 1100 can be a thermal neutron spectrum nuclear fission reactor or a fast neutron spectrum nuclear fission reactor, as desired for a particular application. The temperature detectors detect local temperature in the nuclear fission reactor 1100 and generate a signal 1114 indicative of a detected local temperature. The signal 1114 is transmitted to a control system 1120 in any acceptable manner, such as without limitation, fluid coupling, electrical coupling, optical coupling, radiofrequency transmission, acoustic coupling, magnetic coupling, or the like. Responsive to the signal 1114 indicative of the detected local temperature, the control system 1120 determines an appropriate correction (positive or negative) to local neutronic reactivity in the nuclear fission reactor 1100 to return the nuclear fission reactor 100 to desired operating parameters (such as desired local temperatures for full reactor power). To that end, the control system 1120 generates a control signal 1124 indicative of a desired correction to local neutronic reactivity. The control signal 1124 is transmitted to a dispenser 1130 of neutron absorbing material. The signal 1124 suitably is transmitted in the same manner as the signal 1114. The neutron absorbing material suitably is any neutron absorbing material as desired for a particular application, such as without limitation Li6, B10, or Gd. The dispenser 1130 suitably is any reservoir and dispensing mechanism acceptable for a desired application, and may, for example, have the reservoir located remotely (e.g., outside the neutron reflector of the nuclear fission reactor 1100) from the dispensing mechanism 1130. The dispenser 1130 dispenses the neutron absorbing material within the nuclear fission reactor core responsive to the control signal 1124, thereby altering the local neutronic reactivity. Referring now to FIG. 11B and given by way of non-limiting example, exemplary thermal control may be established with a neutron absorbing fluid. A thermally coupled fluid containing structure 1140 contains a fluid in thermal communication with a local region of the nuclear fission reactor 1100. The fluid in the structure 1140 expands or contracts responsive to local temperature fluctuations. Expansion and/or contraction of the fluid is operatively communicated to a force coupling structure 1150, such as without limitation a piston, located external to the nuclear fission reactor 1100. A resultant force communicated by the force coupling structure 1150 is exerted on neutron absorbing fluid in a neutron absorbing fluid containing structure 1160. The neutron absorbing fluid is dispensed accordingly from the structure 1160, thereby altering the local neutronic reactivity. In another example, a neutron moderating fluid may be used instead of, or in addition to, the neutron absorbing fluid. The neutron moderating fluid changes the neutron energy spectrum and lowers the mean neutron energy of the local neutron environment, thereby driving down or lowering neutronic reactivity of nuclear fission fuel within the nuclear fission reactor 1100. In another example, the neutron absorbing fluid and/or the neutron modifying fluid may have a multiple phase composition (e.g., solid pellets within a liquid). FIG. 11C illustrates details of an exemplary implementation of the arrangement shown in FIG. 11B. Referring now to FIG. 11C, fuel power density in a nuclear fission reactor 1100′ is continuously regulated by the collective action of a distributed set of independently-acting thermostating modules, over very large variations in neutron flux, significant variations in neutron spectrum, large changes in fuel composition and order-of-magnitude changes in power demand on the reactor. This action provides a large negative temperature coefficient of reactivity just above the design-temperature of the nuclear fission reactor 1100′. Located throughout the fuel-charge in the nuclear fission reactor 1100′ in a 3-D lattice (which can form either a uniform or a non-uniform array) whose local spacing is roughly a mean free path of a median-energy-for-fission neutron (or may be reduced for redundancy purposes), each of these modules includes a pair of compartments 1140′ and 1160′, each one of which is fed by a capillary tube. The small thermostat-bulb compartment 1160′ located in the nuclear fission fuel-contains a thermally sensitive material, such as without limitation, Li7, whose neutron absorption cross-section may be low for neutron energies of interest, while the relatively large compartment 1140′ positioned in a different location (e.g., on the wall of a coolant tube) may contain variable amounts of a neutron absorbing material, such as without limitation, Li6, which has a comparatively large neutron absorption cross-section. Lithium melts at 453 K and 1-bar-boils at 1615 K, and therefore is a liquid across typical operating temperature ranges of the nuclear fission reactor 1100′. As the fuel temperature rises, the thermally sensitive material contained in the thermostat-bulb 1160′ expands, and a small fraction of it is expelled (approximately 10−3, for a 100K temperature change in Li7), potentially under kilobar pressure, into the capillary tube which terminates on the bottom of a cylinder-and-piston assembly 1150′ located remotely (e.g., outside of the radiation shield) and physically lower than the neutron absorbing material's intra-core compartment 1140′ (in the event that gravitational forces are to be utilized). There the modest volume of high-pressure thermally sensitive material drives a swept-volume-multiplying piston in the assembly 1150′ which pushes a potentially three order-of-magnitude larger volume of neutron absorbing material through a core-threading capillary tube into an intra-core compartment proximate to the thermostat-bulb which is driving the flow. There the neutron absorbing material, whose spatial configuration is immaterial as long as its smallest dimension is less than a neutron mean free path, acts to absorptively depress the local neutron flux, thereby reducing the local fuel power density. When the local fuel temperature drops, neutron absorbing material returns to the cylinder-and-piston assembly 150′ (e.g., under action of a gravitational pressure-head), thereby returning the thermally sensitive material to the thermostat-bulb 1160′ whose now-lower thermomechanical pressure permits it to be received. It will be appreciated that operation of thermostating modules does not rely upon the specific fluids (Li6 and Li7) discussed in the above exemplary implementation. In one exemplary embodiment, the thermally sensitive material may be chemically, not just isotropically, different from the neutron absorbing material. In another exemplary embodiment, the thermally sensitive material may be isotropically the same as the neutron absorbing material, with the differential neutron absorbing properties due to a difference in volume of neutronically exposed material, not a difference in material composition. Referring now to FIG. 12, in another embodiment a propagating nuclear fission deflagration wave reactor 1200 operates at core temperatures significantly lower than core temperatures of nuclear fission reactors of other embodiments. While nuclear fission reactors of other embodiments may operate at core temperatures in the order of around 1,000K or so, (e.g., to enhance electrical power conversion efficiency) the propagating nuclear fission deflagration wave reactor 1200 operates at core temperatures of less than around 550K, and some embodiments operate at core temperatures of between around 400K and around 500K. Reactor coolant 1210 transfers heat from nuclear fission in the propagating nuclear fission deflagration wave reactor 1200. In turn thermal energy 1220 is transferred from the reactor coolant 1210 to a thermally driven application. Given by way of non-limiting examples, exemplary thermally driven applications include desalinating seawater, processing biomass into ethanol, space-heating, and the like. In another embodiment, a propagating nuclear fission deflagration wave reactor 1200 may operate at core temperatures above 550K, and utilize thermal energy 1220 from the reactor coolant 1210 for thermally driven applications instead of, or in addition to, electrical power generation applications. Given by way of non-limiting examples, exemplary thermally driven applications include thermolysis of water, thermal hydrocarbon processing, and the like. Referring now to FIG. 13, in another embodiment nuclear fission fuel can be removed after it has been burned. A nuclear fission deflagration wave propagating burnfront 1310 is initiated and propagated in a modular nuclear fission reactor core 1300 along a direction indicated by arrows 1320 toward modules 1340 of nuclear fission fuel material, thereby establishing a region 1330 of maximized reactivity as discussed above. As discussed above, the modules 1340 of nuclear fission fuel material may be considered “burnt” after the propagating burnfront 1310 has propagated the region 1330 of maximized reactivity through the module 1340 of nuclear fission fuel material. That is, the modules 1340 of nuclear fission fuel material “behind” the region 1330 of maximized reactivity may be considered “burnt”. Any desired number of the “burnt” modules 1340 of nuclear fission fuel material (behind the region 1330 of maximized reactivity) are removed, as generally indicated at 1350. As generally indicated at 1360, nuclear fission fuel material has been removed from the nuclear fission reactor core 1300. Referring now to FIGS. 14A and 14B, according to other embodiments nuclear fission fuel can be re-burned in place without reprocessing. As shown in FIG. 14A, a propagating nuclear fission deflagration wave reactor 1400 includes regions 1410 and 1420. A nuclear fission deflagration wave burnfront 1430 is initiated and propagated through the region 1410 toward the region 1420. The nuclear fission deflagration wave burnfront 1430 propagates through the region 1420 as a nuclear fission deflagration wave burnfront 1440. After the nuclear fission deflagration wave burnfront 1440 propagates into region 1420, and either before or after it reaches an end of the propagating nuclear fission deflagration wave reactor 1400, the nuclear fission deflagration wave burnfront 1440 is redirected or re-initiated and retraces a path of propagation away from the end of the propagating nuclear fission deflagration wave reactor 1400 back toward the region 1410. The nuclear fission deflagration wave burnfront 1440 propagates through the region 1410 as a nuclear fission deflagration wave burnfront 1450 away from the region 1420 toward an end of the propagating nuclear fission deflagration wave reactor 1400. The nuclear fission fuel in regions 1410 and 1420 is different during the repropagation of nuclear fission deflagration wave burnfronts 1440 and 1450 than it was during the previous propagation of nuclear fission deflagration wave burnfronts 1430 and 1440, due to changes in the amounts of fissile isotopes and the amounts of fission product “ash”. The neutron environment may differ during propagation and repropagation due to the above differences in the nuclear fission fuel, as well as other factors, such as without limitation, possible changes in the control of neutron modifying structures, thermal heat extraction levels, or the like. As shown in FIG. 14B (and as briefly mentioned in reference to FIG. 3C), the geometry of an embodiment of the propagating nuclear fission deflagration wave reactor 1400 forms a closed loop, such as an approximately toroidal shape. In this exemplary embodiment, the propagating nuclear fission deflagration wave reactor 1400 includes the regions 1410 and 1420 and a third region 1460 different from the regions 1410 and 1420. The nuclear fission deflagration wave burnfront 1430 is initiated and propagated through the region 1410 toward the region 1420. The nuclear fission deflagration wave burnfront 1430 propagates through the region 1420 as the nuclear fission deflagration wave burnfront 1440. The nuclear fission deflagration wave burnfront 1440 propagates through the region 1460 as a nuclear fission deflagration wave burnfront 1470. When the nuclear fission deflagration wave burnfronts 1430, 1440, and 1470 have propagated completely through the regions 1410, 1420, and 1460, respectively, nuclear fission fuel material in the regions 1410, 1420, and 1460 can be considered “burnt”. After the nuclear fission fuel material has been burnt, the nuclear fission deflagration wave burnfront 1430 is re-initiated and propagates through the region 1410 as a nuclear fission deflagration wave burnfront 1450. The re-initiation in region 1410 may occur without limitation, through the action of a nuclear fission igniter, such as discussed earlier, or may occur as a result of the decay and/or removal of nuclear fission products from the nuclear fission fuel material in region 1410, or may occur as the result of other sources of neutrons or fissile material, or may occur due to control of neutron modifying structures, as discussed previously. In another exemplary embodiment, the nuclear fission deflagration wave may potentially propagate in a plurality of directions. One or more propagation paths may be established, and may thereafter split into one or more separate propagation paths. The splitting of propagation paths may be accomplished without limitation by such methods as the configuration of the nuclear fission fuel material, the action of neutron modifying structures as discussed earlier, or the like. Propagation paths may be distinct, or may be reentrant. Nuclear fission fuel material may be burnt once, never, or multiple times. Repropagation of a nuclear fission deflagration wave multiple times through a region of nuclear fission fuel material may involve either the same or a different propagation direction. While some of the embodiments described previously illustrate nuclear fission fuel cores of substantially constant chemical and/or isotropic materials, in some approaches nuclear fission fuel cores of nonuniform material may be used. For example, in some approaches nuclear fission fuel cores may include regions having different percentages of uranium and thorium. In other approaches, nuclear fission fuel cores may include regions of different actinide or transuranic isotopes, such as without limitation different isotopes of thorium or different isotopes of uranium. In addition, mixtures of such different combinations may also be appropriate. For example, mixtures of thorium and of different uranium isotope ratios may provide different burning rates, temperatures, propagation features, localization, or other features. In other approaches, the nuclear fission fuel cores may include mixtures of “breedable” isotopes (such as Th232 or U238) along with other fissionable actinide or transuranic elements, such as without limitation, uranium, plutonium, americium, or the like. Additionally, such variations in chemicals, isotopes, cross sections, densities, or other aspects of the fuel or may vary radially, axially or in a variety of other spatial manners. For example, such variations may be defined according to anticipated variations in energy demand, aging, or other anticipated variations. In one aspect, where growth of energy demand in a region would be reasonably anticipated, it may be useful to define the fuel or materials to correlate to an expected increased demand of the region. In still another aspect, such variations may be implemented according to other approaches described herein. For example, the variations may be defined after initiation of burning using the modular approach is described herein or the multipath approaches described herein. In other approaches, movement of portions of the material may produce the appropriate material concentrations, positioning, ratios, or other characteristics. While the embodiments above have illustrated propagating nuclear fission deflagration wavefronts in fixed or variable fuel cores, in one aspect, propagating nuclear fission deflagration wavefronts may remain substantially spatially fixed while the fuel core or portions of the fuel core move relative to the wavefront. In one such approach, movement of the nuclear fission fuel core to maintain substantially localized positioning of the propagating nuclear fission deflagration wavefront can stabilize, optimize, or otherwise control thermal coupling to a cooling or heat transfer system. Or, in another aspect, controlled positioning of the propagating nuclear fission deflagration wavefront by physically displacing the nuclear fission fuel can simplify or reduce constraints upon other aspects of the nuclear fission reactor, such as the cooling system, neutron shielding, or other aspects of neutron density control. While a number of exemplary embodiments and aspects have been illustrated and discussed above, those of skill in the art will recognize certain modifications, permutations, additions, and sub-combinations thereof. It is therefore intended that the following appended claims and claims hereafter introduced are interpreted to include all such modifications, permutations, additions, and sub-combinations as are within their true spirit and scope.
	claims	1. An apparatus comprising:a first electrode having a substantially cylindrical inner surface that has a longitudinal axis and forms at least a portion of a confining wall, wherein the confining wall at least partially encloses a confinement region;a second electrode located within a region interior to the first electrode and separated from the first electrode by at least the confinement region;at least one magnet configured to provide a magnetic field through the confinement region, at least a portion of the magnetic field in the confinement region being substantially parallel to the longitudinal axis;an inlet to the confinement region for permitting introduction of a fluid to the confinement region, the fluid containing a first reactant;a second reactant; anda control system comprising one or both of a voltage source and a current source and configured to (a) control a potential of an electric field substantially orthogonal to the longitudinal axis, the potential between the first electrode and the second electrode being sufficient to produce an electrical current from the first electrode to the second electrode; (b) generate, from the first reactant, a weakly ionized plasma of ions and neutrals; and (c) by interaction of the electric field and the magnetic field, produce a Lorentzian force that induces azimuthal rotation of the ions around the longitudinal axis, the azimuthal rotation of the ions imparting azimuthal rotation to neutrals of the first reactant, and promoting repeated collisions between one or both of the ions and the neutrals with the second reactant; wherein, during operation:the repeated collisions between the neutral particles and the reactant produce an interaction produces a product having a nuclear mass that is different from a nuclear mass of any of the nuclei of the neutrals and the second reactant and,a mole fraction of the ions to the neutrals in the weakly ionized plasma is in the range about of 0.0001% to about 1%. 2. The apparatus of claim 1, wherein the second electrode has a diameter of at most about 0.5 inches. 3. The apparatus of claim 1, wherein the second electrode has a length in a direction parallel to the longitudinal axis. 4. The apparatus of claim 1, further comprising a ceramic block thermally coupled to the second electrode and configured to remove thermal energy from the second electrode via conduction. 5. The apparatus of claim 1, further comprising a linear actuator that is configured to move the second electrode in a direction that is substantially parallel to the longitudinal axis. 6. The apparatus of claim 1, wherein the first electrode and the second electrode are separated by a gap of between about 0.1 mm and about 20 cm. 7. The apparatus of claim 1, wherein the at least one magnet comprises at least one permanent magnet having opposite magnetic poles offset from one another in the direction of the longitudinal axis. 8. The apparatus of claim 1, wherein the at least one magnet comprises two permanent magnets separated from one another by at least the confinement region and in the direction of the longitudinal axis. 9. The apparatus of claim 1, wherein the second electrode is coated with an electron emitting material. 10. The apparatus of claim 1, wherein the reactant comprises boron 11. 11. The apparatus of claim 1, wherein the confining wall comprises one or both of a refractory metal and a stainless steel. 12. The apparatus of claim 1, wherein the interaction is a fusion reaction. 13. The apparatus of claim 1, wherein the interaction is an aneutronic fusion reaction. 14. The apparatus of claim 1, wherein the neutral particles comprise one or more of neutral hydrogen, deuterium, and tritium. 15. The apparatus of claim 1, wherein, during operation, the neutral particles in the confinement region proximate the second electrode have a concentration of at least about 1020/cm3. 16. The apparatus of claim 1, further comprising one or more electron emitters configured to, during operation, emit electrons into an electron-rich region adjacent to the second electrode. 17. The apparatus of claim 1, wherein the confinement region proximate the second electrode comprises an electron-rich region having at least about 106/cm3 more electrons than positively charged particles. 18. The apparatus of claim 16 wherein, during operation, the electron-rich region includes an electric field strength of at least about 106 V/m.
	claims	1. A rupture disk for a device for protecting against overpressure inside an apparatus, the disk consisting of a part of generally circular form comprising two planar faces substantially parallel to one another, and two scores each located on a circumference, wherein the circumferences of the two scores are different from one another, the score located on the larger circumference being produced on one of the planar faces, called bottom face, whereas the score located on the smaller circumference is produced on the other of the planar faces, called top face, the score located on the top face further being configured to rupture at a first pressure P0 under a first gas whereas the score located on the bottom face being configured to rupture at a second pressure P1 different from P0 under a second gas. 2. The rupture disk as claimed in claim 1, one and/or the other of the scores being continuous over its circumference. 3. The rupture disk as claimed in claim 1, one and/or the other of the scores being discontinuous on its circumference. 4. The rupture disk as claimed in claim 1, wherein the circumferences of the two scores are concentric to one another. 5. The rupture disk as claimed in claim 1, wherein the circumferences of the two scores are centered on the center of the circular part. 6. The rupture disk as claimed in claim 1, wherein the part is made of a steel chosen from the ferritic, martensitic, ferrito-bainitic, bainitic, ferrito martensitic, ferrito-perlitic, and perlitic steels. 7. The rupture disk as claimed in claim 1, wherein the part has a thickness of 1 to 5 mm. 8. The rupture disk as claimed in claim 1, wherein the depth of the scores is less than or equal to a value corresponding to approximately 70% of the thickness of the part. 9. The rupture disk as claimed in claim 1, wherein the radius at the bottom of the scores is between 0.1 and 0.6 mm. 10. The rupture disk as claimed in claim 1, wherein the aperture angle of the scores is less than or equal to 50°. 11. An apparatus, intended to contain, in succession, two gases of different chemical nature, comprising:a wall in which an aperture is formed;a device for protecting the apparatus against overpressures comprising:a rupture disk as claimed in claim 1, the bottom face of which is in direct contact with the pressure contained in the apparatus,at least one bearing element, designed to press the disk against the wall with the scores of the disk over the aperture. 12. The apparatus as claimed in claim 11, wherein the protection device comprises:sealing means arranged between the periphery of the disk and the wall around the aperture,at least one holding element, called closing cap, bearing against the bearing element, and fixed by tightening to the apparatus so as to hold the bearing element and the rupture disk on the apparatus and ensure the seal at the level of the sealing means. 13. The apparatus as claimed in claim 12, wherein the closing cap and the apparatus is pierced respectively with tapped holes facing one another to house the screws for tightening the closing cap on the apparatus. 14. The apparatus as claimed in claim 11, wherein the bearing element is a ring whose bottom face bears against the top face of the disk and whose aperture is arranged facing the aperture of the wall. 15. The apparatus as claimed in claim 11, further comprising sealing means arranged between the periphery of the ring and the closing cap. 16. The apparatus as claimed in claim 15, the sealing means consisting of one or more O-ring seals. 17. A method comprising containing under pressure, two gases of different chemical nature, by the apparatus as claimed in claim 11. 18. The method as claimed in claim 17, wherein the apparatus is a hydrogen production and storage apparatus.
063255383	claims	1. An isolation system to be used with radiation equipment for protecting operating personnel from stray radiation, said radiation equipment comprising: a) a source of radiation; b) a shield that comprises: c) wherein the shield forms an envelope about an area in between the radiation source and the patient. each trapezoid surface consisting of two overlapping sections, each section consisting of a plurality of segments movably connected together, said segments being connected together by flexible material, said overlapping sections being slidably connected to each other. each of said stabilizing rods comprising a plurality of telescoping rod sections, said stabilizing rods supporting said second shield section having means at one end for connecting said stabilizing rods to said radiation source, and means at another end for connecting said stabilizing rods to said middle section, said stabilizing rods supporting said third shield section having connectors at one end for connecting said stabilizing rods to said camera, and connectors at another end for connecting said stabilizing rods to said middle shielding section. each of said side sections comprising a middle piece and two end pieces, each of said middle pieces comprising a rectangular plate, each having a rectangular opening and a lower and an upper end, each of said lower and upper ends of said rectangular plates being connected to a horizontal bar, each of said horizontal bars having two parallel slots extending there through. said first and second corner pieces being connected together at an angle, a pair of horizontal bars extending from each of said corner pieces, said pair of horizontal bars extending from said first corner piece engaging said slots in said horizontal bars on said middle piece, said pair of horizontal bars extending from said second corner piece engaging a U-shaped bar, said U-shaped bar having a vertical leg and two horizontal legs. said horizontal legs on said U-shaped bar entering said hollow horizontal bars and engage said spring. said flexible walls extending across said table, said central apertures of said flexible screens having a slot extending to a periphery of said flexible screens, and means for closing said slot. a) frame that includes a table for supporting a patient; b) a source of radiation; c) a camera positioned to receive radiation that has passed through the patient; d) a shield that comprises: e) wherein the shield forms an envelope about an area in between the radiation source and the camera. 2. The isolation system of claim 1 further comprising a camera and a third shielding section attached to the middle section and the camera. 3. The isolation system of claim 2 wherein the camera and radiation source are generally opposite each other on opposing sides of said middle section. 4. The isolation system of claim 2 wherein one of said second and third sections is tapered. 5. The isolation system of claim 2 wherein both of said second and third sections is tapered. 6. The isolation system of claim 2 wherein one of said second and third sections have a flexible wall. 7. The isolation system of claim 2 wherein both of said second and third sections has a flexible wall. 8. The isolation system of claim 2 wherein one of said second and third sections is in the form of a bellows like structure having folded wall portions. 9. The isolation system of claim 2 wherein both of said second and third sections are in the form of a bellows like structure having folded wall portions. 10. The isolation system as claimed in claim 5, wherein each of said tapered sections comprises four trapezoid side surfaces, 11. The isolation system as claimed in claim 2, wherein each of said shield sections is supported by a plurality of stabilizing rods, 12. The isolation system of claim 11 wherein the connectors include a plurality of connector rings. 13. The isolation system of claim 12 wherein each one of said connector rings has at least two loops, one of said at least two loops being connected to one of said stabilizing rods, and another of said at least two loops being connected to said shield sections. 14. The isolation system as claimed in claim 1, wherein said middle shielding section comprises two side sections, 15. The isolation system as claimed in claim 14, wherein each of the end pieces of said side sections comprise a vertical support having a fist corner piece connected to a second corner piece, 16. The isolation system as claimed in claim 15, wherein said pair of horizontal bars extending from said second corner piece are hollow, and have a spring positioned within said hollow space, 17. The isolation system as claimed in claim 6, wherein flexible walls have a central aperture therein are located between and attached to said vertical legs of said U-shaped bars, 18. The isolation system of claim 1 further comprising a table that supports said middle section. 19. The isolation system of claim 18 further comprising a camera and a third shielding section attached to the middle section and the camera. 20. The isolation system of claim 18 wherein the camera and radiation generator are generally opposite each other on opposing sides of said middle section. 21. The isolation system of claim 18 wherein one of said second and third sections is tapered. 22. The isolation system of claim 18 wherein both of said second and third sections is tapered. 23. The isolation system of claim 18 wherein one of said second and third sections has a flexible wall. 24. The isolation system of claim 18 wherein both of said second and third sections has a flexible wall. 25. The isolation system of claim 18 wherein one of said second and third sections is in the form of a bellows like structure having folded wall portions. 26. The isolation system of claim 18 wherein one of said second and third sections is in the form of a bellows like structure having folded wall portions. 27. An isolation system to be used with radiation equipment for protecting operating personnel from stray radiation, said radiation equipment comprising:
046735470	abstract	The invention relates to apparatus for the separation of hydrogen and/or terium and tritium from an inert gas flow which is contaminated with hydrogen and/or deuterium and/or tritium, wherein the inert gas which is to be purified is conducted along the primary side of an exchange wall for hydrogen isotopes. The secondary side of the exchange wall has applied thereto an agent which so chemically reacts with the permeating hydrogen isotopes that the hydrogen isotope or the hydrogen isotopes which are to be separated out of the inert gas flow which is to be purified, are bound in a reaction product which is transportable in a gas flow which is not capable of permeating through the exchange wall, and wherein along the secondary side of the exchange wall there is conducted a carrier gas flow which conveys off the reaction product.
	claims	1. A plasma confinement system comprising:an inner electrode;an intermediate electrode that at least partially surrounds the inner electrode; andan outer electrode that at least partially surrounds the intermediate electrode, the outer electrode comprising:a solid conductive shell comprising:a solid conductive outer shell; anda solid inner shell that is disposed within the solid conductive outer shell and in contact with the solid conductive outer shell,wherein the solid inner shell comprises:an axial wall that at least partially encircles a longitudinal axis of the plasma confinement system; anda radial wall that couples the axial wall to the solid conductive outer shell,wherein the radial wall and a first end of the outer electrode form a pool region,an electrically conductive material disposed on the solid conductive shell, wherein the electrically conductive material has a melting point within a range of 180° C. to 800° C. at 1 atmosphere of pressure; anda pumping system configured to circulate the electrically conductive material over the outer electrode when the electrically conductive material is in a liquid state. 2. The plasma confinement system of claim 1, further comprising a feeding mechanism that is configured to move the inner electrode along the longitudinal axis of the plasma confinement system. 3. The plasma confinement system of claim 1, further comprising a cooling system that is configured to cool the inner electrode during operation of the plasma confinement system. 4. The plasma confinement system of claim 1, further comprising:a first power supply configured to apply a voltage between the inner electrode and the intermediate electrode; anda second power supply configured to apply a voltage between the inner electrode and the outer electrode. 5. The plasma confinement system of claim 1, wherein the electrically conductive material comprises one or more of lithium, lead, or tin. 6. The plasma confinement system of claim 1, the inner electrode having a first end that is at least partially surrounded by the outer electrode, wherein the plasma confinement system is configured to maintain a Z-pinch plasma between the first end of the inner electrode and the electrically conductive material. 7. The plasma confinement system of claim 1,further comprising:a heat exchanger; anda first port configured to guide the electrically conductive material from the heat exchanger into the pool region. 8. The plasma confinement system of claim 1, wherein the pumping system is configured to circulate the electrically conductive material such that movement of the electrically conductive material includes one or more of an azimuthal component or an axial component with respect to the longitudinal axis of the plasma confinement system. 9. The plasma confinement system of claim 1, a first end of the inner electrode being formed of graphite or carbon fiber. 10. The plasma confinement system of claim 1, further comprising one or more gas ports configured to direct gas into an acceleration region between the inner electrode and the intermediate electrode. 11. The plasma confinement system of claim 1, wherein the plasma confinement system allows current to flow between the inner electrode and the outer electrode through both a Z-pinch plasma and the electrically conductive material. 12. The plasma confinement system of claim 1, the inner electrode having a first end that is at least partially surrounded by the outer electrode. 13. The plasma confinement system of claim 12, wherein the first end of the inner electrode is rounded. 14. The plasma confinement system of claim 1, further comprising an insulator between a second end of the outer electrode and the intermediate electrode. 15. The plasma confinement system of claim 14, wherein the plasma confinement system is configured to maintain a Z-pinch plasma between a first end of the inner electrode and the electrically conductive material on the longitudinal axis of the plasma confinement system. 16. The plasma confinement system of claim 15, wherein the Z-pinch plasma has axial flow that varies in a radial direction. 17. The plasma confinement system of claim 7, the axial wall comprising an end that faces a second end of the outer electrode, the plasma confinement system further comprising:a first pump configured to move the electrically conductive material from the pool region to a region that is outside the axial wall and separated from the pool region by the radial wall. 18. The plasma confinement system of claim 17, the plasma confinement system further comprising:a second pump configured to move the electrically conductive material from the pool region to the region that is outside the axial wall and separated from the pool region by the radial wall. 19. The plasma confinement system of claim 1, wherein the intermediate electrode forms a cavity and the inner electrode is positioned at least partially within the cavity. 20. The plasma confinement system of claim 19, the inner electrode having a first end that is at least partially surrounded by the outer electrode.
	summary
	summary
047073233	claims	1. A closure for an inclined duct having an open upper end and defining a downwardly extending passageway comprising: cap means for sealing engagement with said open upper end of said duct; an array of vertically aligned plug members each having a cross-sectional area substantially conforming to the cross-sectional area of said passageway at least adjacent an upper end thereof and being freely movable therein, each of said plug members comprising a cylinder each cylinder having a longitudinally-extending axle at each end; means interconnecting each of said vertically aligned plug members, said means providing for free movement of said plug members only in the plane in which said duct is inclined, said means comprising a plurality of end plates interconnecting axles of adjacent cylinders; means connecting said cap means to the uppermost plug of said array and; hoist means located directly over the open upper end of said duct for moving said closure between an upper position at which said lowermost plug of said array is above the open upper end of said duct and a lower position at which said cap means is in engagement with the open upper end of said duct. cap means for sealing engagement with said open upper end of said duct; an array of vertically aligned plug members each having a substantially rectangular cross-sectional area conforming to the cross-sectional area of the passageway, at least adjacent an upper end thereof and being freely movable therein, each of said plug members comprising a cylinder, each cylinder having a longitudinally extending axle at each end; means interconnecting each of said vertically aligned plug members, said means providing for free movement of said plug members only in the plane in which said duct is inclined, said means comprising a plurality of end plates interconnecting axles of adjacent cylinders; means connecting said cap means to the uppermost plug of said array; hoist means located directly over the upper end of said duct for moving said closure between an upper position at which said lowermost plug of said array is above the open upper end of said duct, and a lower position at which said cap means is in engagement with the open upper end of said duct; and means associated with said cap means for guiding at least a lower portion of said array of said vertically aligned plug members into the open upper end of said duct. 2. The closure of claim 1 further including latch means associated with said cap means for securing said cap means to the open upper end of said duct. 3. The closure of claim 2 wherein said latch means is attached to said cap means and connected to said hoist means for engaging said latch means when said cap means engages the open end of said duct and for releasing said latch means when said hoist means moves in an upward direction. 4. The closure of claim 1 further including means for guiding at least a lower portion of said array of vertically aligned plug members into the open upper end of said duct. 5. The closure of claim 4 wherein said guide means comprises a vertical member in fixed relation to said duct means and said cap means includes an alignment member in vertically slideable engagement with said vertical member. 6. The closure of claim 5 wherein said vertical member has a lower terminal end located above the open end of said duct. 7. The closure of claim 1 wherein said plug members adjacent a lower portion of said array comprise thermal insulators. 8. The closure of claim 1 wherein said duct is a fuel transfer duct in a nuclear reactor. 9. In a liquid metal cooled fast breeder reactor, a closure for an inclined fuel transfer duct, said duct having an open upper end and defining a downwardly extending substantially rectangular passageway comprising: 10. The closure of claim 9 further including latch means associated with said cap means for securing said cap means to the open end of said duct. 11. The closure of claim 10 wherein said guide means comprises a vertical member in fixed relation to said duct means and said cap means includes an alignment member in vertically slideable engagement with said vertical member. 12. The closure of claim 11 wherein said vertical member has a lower terminal end located at a higher elevation than the open end of said duct. 13. The closure of claim 12 wherein said latch means is attached to to said cap means and connected to said hoist means for engaging said latch means when said cap means engages the open end of said duct, and for releasing said latch means when said hoist means moves in an upward direction.
051749475	description	DETAILED DESCRIPTION OF THE INVENTION In the following description, like reference characters designate like or corresponding parts throughout the several views of the drawings. Also in the following description, it is to be understood that such terms as "forward", "rearward", "left", "right", "upwardly", "downwardly", and the like, are words of convenience and are not to be construed as limiting terms. Referring now to the drawings, and particularly to FIGS. 1 and 2, there is illustrated a nuclear fuel pellet turning apparatus of the present invention, being generally designated 10. The pellet turning apparatus 10 is used by an operator for handling a plurality of nuclear fuel pellets P at a time and facilitating surface inspection of the pellets P (FIGS. 18 and 23). The pellet turning apparatus 10 is disposed upon a work table 12 and under an exhaust hood 14 contained in an enclosed isolation chamber. A pair of spaced upright brackets 16, 18 are mounted on the work table 12 adjacent the pellet turning apparatus 10 for receiving and supporting a pellet supply tray 20 next to the pellet turning apparatus 10. In its basic components, the pellet turning apparatus 10 includes a pellet turner assembly 22 cable of supporting the pellets P and an interface assembly 24 coupled to the pellet turner assembly 22. The pellet turner assembly 22 can be operated to accomplish two functions. First, the pellet turner assembly 22 is operable for producing simultaneous turning of the pellets P in situ (i.e., rolling in place) to permit visual inspection of the entire circumferential surfaces of the pellets. Second, the pellet turner assembly 22 is also operable for producing simultaneous tilting of the pellets P to permit visual inspection of the opposite edges of the pellets. The interface assembly 24 is coupled to a pellet loading end 22A of the pellet turner assembly 22 and is operable for disposing a pellet transfer end 20A of the pellet supply tray 20 at a desired elevation above or below the loading end 22A of the pellet turner assembly 22 to facilitate transfer of uninspected pellets from the pellet supply tray 20 to the pellet turner assembly 22 and of inspected pellets from the pellet turner assembly 22 to the pellet supply tray 20. Referring to FIGS. 3-17, there is illustrated the pellet turner assembly 22 which includes a pellet turner deck 26 and a swivel mechanism 28. The pellet turner deck 26 supports pellets P in side-by-side spaced rows. The swivel mechanism 28 supports the pellet turner deck 26 and, in turn, is operable for swiveling and tilting the pellet turner deck 26 to allow simultaneous tilting of the pellets P and thereby permit an operator to easily visually inspect the opposite edges of the pellets for the presence of cracks and chips. More particularly, referring to FIGS. 3-13, the pellet turner deck 26 includes a base frame 30 mounted on the swivel mechanism 28, a plurality of elongated laterally-spaced rod-like members 32 extending between opposite ends of the base frame 30 and defining channels 34 between the members for receiving and arranging the pellets P in rows. The width of the channels 34 is greater than the diameter of the pellets P, as can be seen in FIG. 18, such that the pellets P are not supported upon the rod-like members 32. Also, means are provided on the base frame 30 for supporting the elongated rod-like members 32 at their opposite ends in spaced relation above the base frame 30. Such means takes the configuration of a support bar 36 attached on the right end and a base bar 38 attached on the left end of a network 40 of interconnecting frame members and a flat plate 42 which also make up the base frame 30. The flat plate 42 rigidly mounts the network 40 and, in turn, is attached on the upper end of the swivel mechanism 28. The swivel mechanism 28 is setup to normally support the deck 26 slightly off horizontal, for example, preferably at a five degrees incline perpendicular to the rod-like members 32. As best seen in FIGS. 8-13, the right ends 32A of the rod-like members 32 are bent ninety degrees and extend within vertical grooves 44 defined in laterally-spaced relation on the exterior of the support bar 36. The left ends 32B of the rod-like members 32 are received through holes 46 in the base bar 38 and biased by fasteners 48 and spring washers 50 to retain the right ends 32A in the grooves 44. As shown more specifically in FIGS. 3, 4 and 13-15, the pellet turner deck 26 also includes a pellet rolling plate 52 and an actuating mechanism 54 coupled to the plate 52. The pellet rolling plate 52 has a flat configuration and is movably supported on opposite ledges 56, 58 defined on the interior sides of the support and base bars 36, 38 of the base frame 30 below the pellet channel-defining rod-like members 32. In such position underlying the rod-like members 32, the rolling plate 52 supports the pellets P (FIG. 18) between the pellet channel-defining members 32. The slightly inclined mounting of the deck 26 ensures that all the pellets P are disposed against the rod-like members 32 on the same side of the channels 34. Such relationship enhances the visual inspection carried out by the operator. The actuating mechanism 54 is mounted on the base frame 30 and coupled to the pellet rolling plate 52 The actuating mechanism 54 is operable to produce oscillatory movement of the plate 52 in a direction generally perpendicular to the channel-defining members 32 for producing rolling or turning of the pellets P in situ on the plate and the channel-defining rod-like members 32. More, particularly, as best seen in FIGS. 16 and 17. The actuating mechanism 54 includes a handle 60 attached on a circular disc 62 rotatably mounted in an opening 64A of a bracket plate 64 which is fixed on and extends outwardly from the base frame 30. The actuating mechanism 54 also includes a cam element 66 secured in the circular disc 62 and extending upwardly through a slot 68 formed in a tab 70 which is fixed on and extends outwardly from the pellet rolling plate 52 above the bracket plate 64. As seen in FIG. 16, when the handle 60 is rotated clockwise from its solid line position to dashed line position, the pellet rolling plate 52 is moved linearly from the inwardly-displaced solid line position to outwardly-displaced dashed line position. Rotation of the handle 60 counterclockwise produced movement of the plate 52 in the opposite direction. The total movement of the pellets P is preferably at least approximately on full revolution of the pellet circumference. Referring to FIGS. 1 and 18, there is illustrated the swivel mechanism 28 having a universal ball joint 72 and a handle 74 movable between locking and unlocking positions. As mentioned above, the swivel mechanism 28 supports the pellet turner deck 26 and, in turn, once the handle 74 has been moved to the unlocking position of FIG. 18 is operable for swiveling and tilting the pellet turner deck 26 to simultaneously place the pellets in a tilted or inclined position. When the deck 26 and pellets P have been placed in the desired tilted position, the handle 74 is moved back to the locked position shown in FIG. 1. The above-described actuating mechanism 54 can be operated to oscillate the plate 52 to cause rolling or turning of the tilted pellets in situ. This permits an operator to easily visually inspect the entire perimeters of the end edges of the pellets for the presence of cracks and chips. After one end of the pellets are inspected, the swivel mechanism 28 is operated to rotate the pellet turner deck 26 through 180.degree. and reverse the tilting of the deck 26 to inspect the opposite end edges of the pellets. Referring now to FIGS. 19-25, there is illustrated the interface assembly 24 mounted on the right loading end 22A of the pellet turner deck 26 of the pellet turner assembly 22. The interface assembly 24 basically includes a tray elevating mechanism 76 and an actuating cam mechanism 78 operable to cause raising and lowering of the elevating mechanism 76. The tray elevating mechanism includes a plurality of engaging elements 80 mounted for reciprocal vertical movement on the exterior side of the support bar 36. Each engaging element 80 includes a plate 82 having a pair of vertical slots 84 which receive studs 86 extending from the support bar 36 to mount the plate 82 for vertical reciprocal movement. Each engaging element 80 further has an upstanding pin 88 insertable into a hole 90 (FIGS. 2 and 23) in the transfer end 20A of the pellet supply tray 20. The actuating cam mechanism 78 includes a handle 92 and a plurality of links 94 extending from the handle 92 and between and coupled with the engaging elements 80 of the tray elevating mechanism 76 via cam pins 96 on the links 94 extending through inclined slots 98 in the plates 82. The handle 92 is attached to a plate 100 which, in turn, is pivotally mounted to the base frame 30 by a stud 102. The pivotal movement of the plate 100 produced by pivotal movement of the handle 92 is transmitted and converted to linear movement of the interconnected links 94 via a cam pin 104 which is attached to the leftmost one of the links 94 and extends through an arcuate slot 106 formed through the plate 100. The offset of the radius of the arcuate slot 106 from the pivot point of the plate 100 produces linear movement of the links 94 upon rotation of the plate 100 and handle 92. In such manner, the actuating cam mechanism 78 is coupled to the tray elevating mechanism 76 and operable for moving the elevating mechanism to position the pellet transfer end of the supply tray either above or below the loading end 22A of the pellet turner assembly 22 depending upon the direction of rotation of the handle 92. Positioning of the pellet transfer end 20 A of the pellet supply tray 20 above or below the loading end 22A of the pellet turner assembly 22 correspondingly facilitates transferring of pellets to or from the pellet turner assembly 22. Referring to FIGS. 26-28, there is illustrated a pellet rake 108 which can be employed by an operator with the pellet turning apparatus 10 to slidably move pellets onto and from the deck 26. The rack 108 has a handle 110 at one end, a row of teeth 112 spaced apart by a distance adapting them to fit within the channels 34 of the deck 26, and interconnected by a flat connecting portion 114. The teeth 112 extend at a right angle to the flat portion 114. Referring to FIGS. 29 to 34, there are diagrammatic views of the operations involved in using the pellet turning apparatus 10 of the present invention for inspecting the entire cylindrical surfaces of the pellets (P) for dimensional quality, longitudinal cracks and chipped edges. FIG. 29 shows a pellet supply tray 20 disposed adjacent the pellet turner deck 26 of the turner assembly 22. The deck 26 is locked in stationary position and the transfer end 20A of the tray 20 is coupled to the interface assembly 24 (FIG. 19). The interface assembly 24 is operated to elevate the transfer end 20A of the tray 20 slightly above the loading end 22A of the deck 26. The two rectangles with diagonal lines represent multiple adjacent rows of pellets supported on the supply tray 20. Using the rake 108 (FIG. 26), an operator sweeps a first group of pellets from the supply tray onto the deck 26 to the position shown in FIG. 30 and then sweeps a second group of pellets onto the deck 26 to the position shown in FIG. 31. In order to provide clearance for swiveling and tilting the deck 26, as seen in FIG. 32 the emptied tray 20 is shifted to a position away from the deck 26. Next, as represented by FIG. 33, visual inspection commences with the operator moving the plate 52 (FIG. 15) to turn the pellets a full revolution for cylindrical surface inspection. The turning is accomplished by the operator manually swinging the handle 60 of the actuating mechanism 54 (FIG. 16) coupled to the plate 52 through 180.degree.. Then, edge chip visual is performed by the operator utilizing the swivel mechanism 28 to tilt the deck 26 and by again manually swinging the handle 60 to oscillate the plate 52 and turn the pellets. Makeup pellets are inserted by the operator to replace rejected pellets removed by the operator. Finally, the deck 26 is repositioned to its original orientation and the supply tray 20 is returned to the interfaced relationship with the deck as shown in FIG. 34. However, this time the interface assembly 24 is operated to lower the transfer end 20A of the supply tray 20 slightly below the load end 22A of the deck to facilitate transfer of the inspected pellets from the deck 26 to the supply tray 20. The rake 108 is again used by the operator to sweep the groups of inspected pellets from the deck 26 back onto the supply tray 20. It is thought that the present invention and many of its attendant advantages will be understood from the foregoing description and it will be apparent that various changes may be made in the form, construction and arrangement thereof without departing from the spirit and scope of the invention or sacrificing all of its material advantages, the form hereinbefore described being merely a preferred or exemplary embodiment thereof.
	description	The present invention relates to an apparatus and a method for treating a radioactive nitrate waste liquid capable of reducing nitrate contained in waste water. A nitrate waste liquid generated from nuclear facilities such as a reprocessing plant has high concentration (salt concentration of equal to or more than 1%) as well as being radioactive. Therefore, the nitrate waste liquid cannot be discharged as it is, and is finally cast into a cement-solidified form and disposed underground. In recent years, there has been a concern that when the cement-solidified form contains nitrate, the nitrate may leak and pollute surrounding ground water and soil environment and thus dissolution of the nitrate is studied. There are various methods for treating nitrate such as an electric reduction method, a chemical reduction method, and a biological reduction method. The electric reduction method has problems such as inhibition by heavy metals and generation of ammonia. Besides, the chemical reduction method has problems such as exothermic reaction and generation of ammonia. On the other hand, the biological reduction method allows treatment at an ordinary temperature and pressure and there is no generation of ammonia. Accordingly, the method of treating nitrate by using the biological treatment has been studied (Patent Document 1). However, in the method of Patent Document 1, only one type of carbon source is used, so the method has problems such that the amount of redundant sludge generated with the nitrate reduction treatment is increased and a secondary waste disposal expense is increased. FIG. 11 is a schematic diagram of a configuration of a conventional apparatus for treating a radioactive nitrate waste liquid utilizing an organism. As shown in FIG. 11, a conventional apparatus 100 for treating a radioactive nitrate waste liquid includes a denitrification tank 102 that reduces nitrate, which is present in a nitrate waste liquid 101 generated from nuclear facilities (not shown), to nitrogen gas, a reaeration tank 104 that aerates and mixes a denitrified liquid 103 obtained by denitrification with active sludge, and a precipitation tank 108 that separates a reaerated liquid 105 that is a denitrification-treated liquid discharged from the reaeration tank 104 into precipitated sludge 106 and a treated liquid 107. The denitrification tank 102 includes active sludge containing a large amount of denitrifying bacteria (not shown). In the denitrification tank 102, nitrate ions in a nitrate waste liquid are reduced to nitrogen gas (N2) according to a reaction based on the following formula (1) by an action of an anaerobic microorganism (denitrifying bacteria) and is reduced from the nitrate waste liquid. At this time, a carbon source 121 such as methanol and a pH adjuster 120 are supplied to the denitrification tank 102. Further, a mixer 110 is used to mix inside the denitrification tank 102.NO3−+5/6CH3OH→1/2N2+5/6CO2+7/6H2O+OH− (1) Thereafter, the reaerated liquid 105 passes through the precipitation tank 108 from the reaeration tank 104, and is sent as a treated liquid 107 to a subsequent process (not shown). The sludge 106 precipitated in the precipitation tank 108 is recovered as redundant sludge 131 by a circulating pump 111. Further, a part of the sludge 106 precipitated in the precipitation tank 108 is returned to the denitrification tank 102 via a returned-sludge supplying line 112 and is reused. The sludge not reused is removed from a system as the redundant sludge 131 and is sent to a disposing process (not shown). Further, there is used a method and an apparatus for treating a nitrate-containing waste liquid. The method includes a microbial treatment process of reducing nitrate nitrogen and nitrite nitrogen to nitrogen by using anaerobic denitrifying bacteria in microorganism-containing sludge, and the apparatus includes a plurality of treatment tanks (Patent Document 2). Patent Document 1: Japanese Patent No. 3697037 Patent Document 2: Japanese Patent Application Laid-open No. 2007-105627 However, in the conventional apparatus 100 for treating a radioactive nitrate waste liquid, when the concentration of nitrate in the nitrate waste liquid 101 is low (salt concentration is less than 1%), the denitrifying bacteria is not perished; however, when the concentration of nitrate in the nitrate waste liquid 101 is high (for example, about 1 to 7%), there is a problem that the denitrifying bacteria may be perished due to a rise in pH with the biological denitrification reaction. Further, there is also the following problem. That is, when the concentration of nitrate in the nitrate waste liquid 101 is high (for example, about 1 to 7%), water seeps from a living organism due to a rise in an osmotic pressure, its biology cannot be maintained and the living organism is perished. Therefore, a minute sludge flock is generated and the sludge 106 does not precipitate in the precipitation tank 108, resulting in the sludge 106 flowing backwards with the treated liquid 107. Still another problem is that, when the carbon source 121 supplied in the denitrification tank 102 (for example, organic acid such as acetic acid or sweetener) is changed, there occurs a multiplication of microorganisms, an increase in the generated amount of redundant sludge 131, and an increase in the amount of disposal. There is also a problem for nuclear facilities that the increased generated amount of the redundant sludge 131 leads to huge expenses for disposal, because a secondary waste is disposed underground as a solidified form. That is, there is a problem that the amount of the secondary waste needs to be decreased. Facilities that treat a radioactive nitrate waste liquid need to be installed within a controlled area for radiation, and thus there is a demand for making the treatment facilities compact. In addition, there is a demand that the amount of secondary waste needs to be reduced. Furthermore, as a microbial treatment system to be installed within a controlled area for radiation, there is a demand that it is needed to stabilize reactions and performances. In view of the above problems, an object of the present invention is to provide a method and an apparatus for treating a radioactive nitrate waste liquid capable of efficiently microbially treating a waste liquid having a high nitrate concentration, and the apparatus can be installed within a controlled area for radiation. According to an aspect of the present invention, an apparatus for treating a radioactive nitrate waste liquid, includes: a denitrification tank that accommodates active sludge which adsorbs or takes in a radioactive substance in a nitrate waste liquid containing nitrate and the radioactive substance and in which an anaerobic microorganism that reduces the nitrate to nitrogen gas grows; and a reaeration tank that aerates and mixes a denitrification-treated liquid treated in the denitrification tank with active sludge in which the aerobic microorganism grows. The denitrification tank includes: a pH adjusting unit that supplies a pH adjuster used for adjusting pH of the nitrate waste liquid; a carbon-source supplying unit that supplies a carbon source to the denitrification tank; a first solid-liquid separating unit that separates a denitrified liquid treated with the active sludge into a sludge-containing solid content and a denitrification-treated liquid; and a gas supplying unit that is arranged on a lower side of the first solid-liquid separating unit and supplies gas not containing oxygen into the denitrification tank. The reaeration tank includes: a second solid-liquid separating unit that further separates the denitrification-treated liquid treated with the active sludge into a sludge-containing solid content and a treated liquid; and an air supplying unit that is arranged on a lower side of the second solid-liquid separating unit and supplies air into the reaeration tank. Advantageously, in the apparatus for treating a radioactive nitrate waste liquid, pH of the denitrified liquid in the denitrification tank is from 7.0 to 10.0. Advantageously, in the apparatus for treating a radioactive nitrate waste liquid, the denitrification tank includes a gas circulating line through which nitrogen gas and carbon dioxide gas produced by a reaction between the anaerobic microorganism and the nitrate in the denitrified liquid are circulated into the denitrification tank. Advantageously, the apparatus for treating a radioactive nitrate waste liquid, further includes: a pH sensor that measures pH of the denitrified liquid; and a carbon-dioxide-gas supply-amount adjusting valve that adjusts an amount of carbon dioxide gas supplied to the denitrification tank. Advantageously, in the apparatus for treating a radioactive nitrate waste liquid, a carbon source supplied to the denitrification tank is acetic acid, and the apparatus further includes: a pH sensor that measures pH of the denitrified liquid; and a carbon-source supply-amount adjusting valve that adjusts an amount of a carbon source supplied to the denitrification tank. Advantageously, in the apparatus for treating a radioactive nitrate waste liquid, a carbon source supplied to the denitrification tank is one of or both of organic acid and sweetener, and air is temporarily supplied to the gas circulating line via an air-supply-amount adjusting valve. Advantageously, in the apparatus for treating a radioactive nitrate waste liquid, the denitrification tank includes an oxidation-reduction potential meter that measures an oxidation-reduction potential of the denitrified liquid. Advantageously, in the apparatus for treating a radioactive nitrate waste liquid includes an adjusting tank at an upstream of the denitrification tank. The adjusting tank includes: an electrical conductivity meter that measures a level of electric conductivity; an industrial-water introducing line; and an industrial-water supply-amount adjusting valve that is interposed through the industrial-water introducing line and adjusts an amount of water supplied based on a measurement value of the electrical conductivity meter. Advantageously, in the apparatus for treating a radioactive nitrate waste liquid, the first solid-liquid separating unit and the second solid-liquid separating unit are solid-liquid separating films. According to another aspect of the present invention, an apparatus for treating a radioactive nitrate waste liquid, includes: a denitrification tank that accommodates active sludge which adsorbs or takes in a radioactive substance in a nitrate waste liquid containing nitrate and the radioactive substance and in which an anaerobic microorganism that reduces the nitrate to nitrogen gas grows; and a reaeration tank that aerates and mixes a denitrification-treated liquid treated in the denitrification tank with active sludge in which the aerobic microorganism grows. The denitrification tank includes: a pH adjusting unit that supplies a pH adjuster used for adjusting pH of the nitrate waste liquid; a carbon-source supplying unit that supplies a carbon source to the denitrification tank; a weir that prevents flowing-out sludge from leaking out into the denitrification tank. The reaeration tank includes: a second solid-liquid separating unit that further separates the denitrification-treated liquid treated with the active sludge into a sludge-containing solid content and a treated liquid; and an air supplying unit that is arranged on a lower side of the second solid-liquid separating unit and supplies air into the reaeration tank. According to another aspect of the present invention, a method for treating a radioactive nitrate waste liquid, includes: in a denitrification tank that accommodates active sludge that adsorbs or takes in a radioactive substance in a nitrate waste liquid containing nitrate and the radioactive substances and in which an anaerobic microorganism that reduces the nitrate to nitrogen gas grows, a step at which an active sludge treatment is performed by supplying a carbon source while adjusting pH of the nitrate waste liquid, and at a time of separating a denitrified liquid treated with the active sludge by using a first solid-liquid separating unit into a sludge-containing solid content and a denitrification-treated liquid, gas not containing oxygen is supplied thereby to accelerate a process for reducing an anaerobic microorganism and cleanse the active sludge adhered to the solid-liquid separating unit with the gas; and in a reaeration tank that aerates and mixes the denitrification-treated liquid treated in the denitrification tank with the active sludge, a step of further separating the denitrification-treated liquid treated with the active sludge into a sludge-containing solid content and a reaeration-treated liquid by using a second solid-liquid separating unit. The present invention includes a first solid-liquid separating unit that separates a denitrified liquid treated with active sludge in a denitrification tank into a sludge-containing solid content and a denitrification-treated liquid, and a second solid-liquid separating unit that further separates a reaerated liquid treated in a reaeration tank into a sludge-containing solid content and a reaeration-treated liquid. Therefore, minute sludge can be prevented from flowing out with a treated liquid. 10A to 10I apparatus for treating radioactive nitrate waste liquid 11 nitrate waste liquid 12A to 12G, 12I denitrification tank 14 reaeration tank 21 pH adjuster 22 carbon source 23 denitrified liquid 24 denitrification-treated liquid 25 first solid-liquid separating film 26A, 26B solid content (redundant sludge) 27 reaeration-treated liquid 28 second solid-liquid separating film 29 reaerated liquid 30 diffusion tube 31A gas discharge line 31B gas circulating line 32A, 32B blower 33 pressure valve 34 air supplying unit 35 gas-introducing line 36 pH sensor 37 carbon-dioxide-gas supply-amount adjusting valve 39 carbon-source supply-amount adjusting valve 41 air-supplying line 42 air-supply-amount adjusting valve 43 oxidation-reduction potential meter 50 nitrate waste liquid (undiluted) 51 adjusting tank 52 electrical conductivity meter (EC meter) 53 industrial-water introducing line 54 industrial-water supply-amount adjusting valve 55 nitrate-waste-liquid supplying line 61 industrial water 62 diluter 63 biological treatment device 65 sludge 67 sludge dewatering device 68 dewatered sludge 69 incinerator 70 treated liquid 71 incinerated ashes 72 cleansing water 80 weir 81 stirring unit P3 nitrate-waste-liquid supplying pump Exemplary embodiments of the present invention will be explained below in detail with reference to the accompanying drawings. Note that the present invention is not limited to the embodiments. In addition, constituent elements in the embodiments include those that can be easily assumed by those skilled in the art or that are substantially equivalent. An apparatus for treating a radioactive nitrate waste liquid according to an embodiment of the present invention is explained with reference to the drawings. FIG. 1 is a conceptual diagram of an apparatus for treating a radioactive nitrate waste liquid according to a first embodiment of the present invention. As shown in FIG. 1, an apparatus 10A for treating a radioactive nitrate waste liquid according to the present embodiment is an apparatus for treating a radioactive nitrate waste liquid including a denitrification tank 12A which accommodates active sludge that adsorbs or takes in a radioactive substance in a nitrate waste liquid 11 containing nitrate and the radioactive substance and in which an anaerobic microorganism that reduces the nitrate to nitrogen gas grows, and a reaeration tank 14 in which a denitrification-treated liquid 24 treated in the denitrification tank 12A is aerated and mixed with the active sludge. The denitrification tank 12A further includes a pH adjusting unit (not shown) that supplies a pH adjuster 21 used for adjusting pH of the nitrate waste liquid 11, a carbon-source supplying unit (not shown) that supplies a carbon source 22 to the denitrification tank 12A, a first solid-liquid separating film 25 that is a first solid-liquid separating unit that separates a denitrified liquid 23 treated with the active sludge into a solid content containing sludge (redundant sludge 26A) and a treated liquid, and a gas supplying unit 30 that is arranged on a lower side of the first solid-liquid separating film 25 and supplies gas not containing oxygen (for example, one or both of nitrogen gas (N2) and carbon dioxide gas (CO2) into the denitrification tank 12A. The reaeration tank 14 includes a second solid-liquid separating film 28 or a second solid-liquid separating unit that reaerates the denitrification-treated liquid 24 treated with the active sludge to further separate a reaerated liquid 29 into redundant sludge 26B and a reaeration-treated liquid 27, and an air supplying unit 34 that is arranged on a lower side of the second solid-liquid separating film 28 and uses a blower 32A to supply air into the reaeration tank 14. At a lower portion of the first solid-liquid separating film 25 and the second solid-liquid separating film 28, the gas supplying unit 30 and the air supplying unit 34 are arranged, respectively. When the supplied gas rises, a liquid mixed with the active sludge passes through surfaces of the separating films, and at this time, the liquid only permeates the films, and a solid-liquid separation is performed in this way. The active sludge that is left after the separation adheres to the film surface; however, the film surface is always cleansed with a flow of the supplied gas. As a result, the active sludge that adheres to the film surface is cleansed and reduced. This makes it possible to perform a solid-liquid separation by using the film surface that is always clean. In FIG. 1, reference character P1 denotes a denitrification-treated-liquid supplying pump that supplies the denitrification-treated liquid 24 to the reaeration tank 14 and reference character P2 denotes a reaeration-treated-liquid supplying pump that sends the reaeration-treated liquid 27. In the apparatus 10A for treating a radioactive nitrate waste liquid according to the present embodiment, the first solid-liquid separating film 25 and the second solid-liquid separating film 28 are respectively arranged in the denitrification tank 12A and the reaeration tank 14. Accordingly, the minute sludge can be completely separated into the denitrified liquid 23 or the reaerated liquid 29. This can prevent the sludge from flowing out. Although not limited thereto, examples of types of the first solid-liquid separating film 25 and the second solid-liquid separating film 28 include well-known solid-liquid separating films such as a flat film and a hollow fiber film. As a result, when the minute sludge is completely separated from the denitrification-treated liquid 24 or the reaeration-treated liquid 27, it is possible to decrease the concentration of suspended solid (SS) of the discharged reaeration-treated liquid 27. When the first solid-liquid separating film 25 is installed within the denitrification tank 12A, it is possible to prevent denitrifying bacteria from flowing out, and there is no contamination of any other bacteria. Thus, only a large amount of denitrifying bacteria multiplied within the denitrification tank 12A exist in the tank. As a result, a sufficient amount of sludge can be secured, and the sludge concentration within the denitrification tank 12A can be maintained at a high level. It is possible to maintain a high level of sludge concentration within the denitrification tank 12A, and thus the level of the denitrification performance of the denitrification tank 12A can be enhanced and the apparatus can be made compact. When the first solid-liquid separating film 25 and the second solid-liquid separating film 28 are arranged, a particulate radioactive substance contained in the nitrate waste liquid 11 can be also separated from the reaeration liquid 29. Thus, it is possible to decrease the radioactive concentration of the reaeration-treated liquid 27 discharged to outside and to decrease the radioactive concentration of the reaeration-treated liquid 27. In the apparatus 10A for treating a radioactive nitrate waste liquid according to the present embodiment, the pH of the denitrified liquid 23 in the denitrification tank 12A is preferably adjusted from 7.0 to 10.0. It is preferable that the pH is from 8.0 to 9.5, and more preferably the pH is 8.0 to 9.0. This is because when the pH of the denitrified liquid 23 exceeds 10.0, the microorganism is perished. Moreover, when the pH is less than 7.0, a reaction rate for a treatment for reducing the nitrogen gas in the microorganism is decreased. In the apparatus 10A for treating a radioactive nitrate waste liquid according to the present embodiment, to improve an execution of the treatment for reducing the nitrogen gas by the anaerobic microorganism, the gas supplying unit 30 is arranged on a lower side (bottom surface side of the tank) of the first solid-liquid separating film 25 of the denitrification tank 12A so as to supply gas not containing oxygen (for example, nitrogen gas) to the denitrified liquid 23 of the denitrification tank 12A. By supplying nitrogen gas into the denitrified liquid 23 in the denitrification tank 12A via the gas supplying unit 30, the denitrified liquid 23 in the denitrification tank 12A can be forcedly stirred and the treatment for reducing the nitrogen gas by the microorganisms can be accelerated. Moreover, when nitrogen gas is supplied via the gas supplying unit 30 from below the first solid-liquid separating film 25, the active sludge that adheres to the first solid-liquid separating film 25 can be reduced. As a result, due to a gas cleansing effect, clogging of the film can be prevented. In the apparatus 10A for treating a radioactive nitrate waste liquid according to the present embodiment, nitrogen gas (N2 gas) is supplied by the gas supplying unit. However, the present invention is not limited thereto, and any gas that does not contain oxygen, such as carbon dioxide gas (CO2 gas) and an inactive gas, can be used. Further, in the apparatus 10A for treating a radioactive nitrate waste liquid according to the present embodiment, nitrogen gas (N2) and carbon dioxide gas (CO2 gas) generated in the denitrification tank 12A are released to outside via a gas discharge line 31A. In the apparatus 10A for treating a radioactive nitrate waste liquid according to the present embodiment, initially, the sludge used in sewage sludge or an industrial waste treatment is put in as seed sludge and active sludge multiplied until the active sludge concentration reaches, for example, about 5,000 to 20,000 mg/L, can be used. While the active sludge can be held by a granular carrier or a fibrous carrier, in the exemplary embodiments of the present invention, various types of such carriers are not used, and floating active sludge is preferably used. In the apparatus 10A for treating a radioactive nitrate waste liquid according to the present embodiment, although the present invention is not limited thereto, examples of the carbon source 22 include organic acid such as acetic acid. In the apparatus 10A for treating a radioactive nitrate waste liquid according to the present embodiment, although the present invention is not limited thereto, examples of the pH adjuster 21 include sulfuric acid and hydrochloric acid. In the apparatus 10A for treating a radioactive nitrate waste liquid according to the present embodiment, as the microorganism contained in the active sludge in the denitrification tank 12, a well-known anaerobic microorganism that exhibits a denitrification performance can be used. Although not particularly limited thereto, in the reaeration tank 14, a well-known aerobic microorganism for a reaeration can be used. In the apparatus 10A for treating a radioactive nitrate waste liquid according to the present embodiment, the solid-liquid separating film is used as a solid-liquid separating unit; however, the present invention is not limited thereto, and any unit that can separate the sludge and the liquid can be used. As described above, in the apparatus 10A for treating a radioactive nitrate waste liquid according to the present embodiment, the first solid-liquid separating film 25 and the second solid-liquid separating film 28 are respectively arranged in the denitrification tank 12A and the reaeration tank 14. Accordingly, the minute sludge can be completely separated into the denitrification-treated liquid 24 and the reaeration-treated liquid 27. Thus, the radioactive substance and the sludge can be prevented from flowing out. Moreover, the concentration of suspended solid (SS) in the reaeration-treated liquid 27 that is discharged from the reaeration tank 14 and is separately treated can be improved. Because a sufficient amount of sludge can be secured within the denitrification tank 12A, the sludge concentration within the denitrification tank 12A can be maintained at a high level. Therefore, a high-level denitrification performance of the denitrification tank 12A can be maintained and the apparatus can be made compact. By using the apparatus according to the present embodiment, a denitrification test was conducted on a nitrate waste liquid having an undiluted nitrate nitrogen concentration of 6700 to 9200 mg/L (salt concentration: 4.0 to 5.5%) by using methanol and acetic acid as a carbon source with a test temperature set to 20 to 25° C. (room temperature). When the first solid-liquid separating film was not arranged in the denitrification tank, the denitrification performance reached 2 kg-N/m3/d, whereas when the first solid-liquid separating film was arranged in the denitrification tank, the denitrification performance was improved to 7 kg-N/m3/d. An apparatus for treating a radioactive nitrate waste liquid according to a second embodiment of the present invention is explained with reference to FIG. 2. An apparatus 10B for treating a radioactive nitrate waste liquid according to the present embodiment is configured substantially identical to the apparatus 10A for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and thus like reference letters or numerals are denoted to like constituent elements of the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and redundant explanations thereof will be omitted. FIG. 2 is a schematic diagram of a configuration of the apparatus for treating a radioactive nitrate waste liquid according to the second embodiment of the present invention. As shown in FIG. 2, in the apparatus 10B for treating a radioactive nitrate waste liquid according to the present embodiment, a gas circulating line 31B that circulates within a denitrification tank 12B nitrogen gas (N2) and carbon dioxide gas (CO2 gas) produced by a reaction between the anaerobic microorganism and the nitrate present in the denitrified liquid 23 in the denitrification tank 12B is arranged and branched off from a part of the gas discharge line 31A. As a result, the nitrogen gas (N2) and the carbon dioxide gas (CO2) generated in the denitrification tank 12B are fed to the gas supplying unit 30 by a blower 32B interposed through the gas circulating line 31B and then introduced into the denitrification tank 12B. In this manner, the nitrogen gas (N2) and the carbon dioxide gas (CO2) generated as a result of the microorganism reaction within the denitrification tank 12B are re-circulated and re-used within the denitrification tank 12B. As a result, it is not necessary to separately arrange the nitrogen-gas supplying unit (as shown in FIG. 1) that supplies gas not containing oxygen. As a result, it is possible to eliminate the nitrogen-gas supplying unit and reduce gas purchasing costs. Further, in the apparatus 10B for treating a radioactive nitrate waste liquid according to the present embodiment, there is arranged a pressure valve 33 that discharges a gas to outside from a portion of the gas discharge line 31A that discharges nitrogen gas (N2) and carbon dioxide gas (CO2) generated in the denitrification tank 12B. Thereby, redundant nitrogen gas (N2) and carbon dioxide gas (CO2) can be released by opening the pressure valve 33. Thus, according to the present embodiment, because the nitrogen gas (N2) and the carbon dioxide gas (CO2) generated in the denitrification tank 12B are used, the gas consumes itself efficiently. As a result, the gas supply apparatus and gas purchasing costs can be reduced. An apparatus for treating a radioactive nitrate waste liquid according to a third embodiment of the present invention is explained with reference to FIG. 3. The apparatus for treating a radioactive nitrate waste liquid according to the present embodiment is configured substantially identical to the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and thus like reference numerals are denoted to like constituent elements of the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and redundant explanations thereof will be omitted. FIG. 3 is a schematic diagram of a configuration of the apparatus for treating a radioactive nitrate waste liquid according to the third embodiment of the present invention. As shown in FIG. 3, in an apparatus 10C for treating a radioactive nitrate waste liquid according to the present embodiment, as the gas introduced into a denitrification tank 12C, carbon dioxide gas (CO2) as well as nitrogen gas (N2) is introduced as a pH treating agent. That is, the apparatus 10C for treating a radioactive nitrate waste liquid according to the present embodiment includes a gas-introducing line 35 that introduces carbon dioxide gas (CO2) and nitrogen gas (N2), a pH sensor 36 that measures pH in the denitrified liquid 23, and a carbon-dioxide-gas supply-amount adjusting valve 37 that adjusts the amount of carbon dioxide gas (CO2) supplied as the pH treating agent to the denitrification tank 12C. When the pH of the denitrified liquid 23 measured by the pH sensor 36 is equal to or more than 9.4, the carbon-dioxide-gas supply-amount adjusting valve 37 is opened to lower the pH of the denitrified liquid 23. On the other hand, when the pH in the denitrification tank 12C measured by the pH sensor 36 is equal to or less than 7.5, the carbon-dioxide-gas supply-amount adjusting valve 37 is closed to raise the pH in the denitrification tank 12C. This eliminates necessity of adding a pH adjuster such as sulfuric acid and hydrochloric acid. In this manner, according to the apparatus 10C for treating a radioactive nitrate waste liquid of the present embodiment, when the pH in the denitrification tank 12C is measured by the pH sensor 36 to control the pH in the denitrification tank 12C, the microorganism in the denitrification tank 12C becomes capable of an efficient denitrification reaction, and thus it is possible to prevent inhibition of the denitrification reaction by the microorganism in the denitrification tank 12C. Further, in the apparatus 10C for treating a radioactive nitrate waste liquid according to the present embodiment, carbon dioxide gas (CO2) can be used as the pH adjuster 21. This is because when the hydrochloric acid is used as the pH adjuster 21, the apparatus is eroded, and when the sulfuric acid is used, sodium sulfate is produced. In the latter case, the amount to be mixed has its limit, and accordingly a solidified form increases, resulting in an increase in disposal expense. On the other hand, when the carbon dioxide gas (CO2) is used as the pH adjuster 21, a remaining salt caused when secondary waste produced when the reaeration-treated liquid 27 is separately treated is solidified becomes sodium carbonate or sodium hydrogen carbonate. The result is that it is possible to reduce the amount of solidified form to be generated that is obtained at the time of incinerating and solidifying a concentrate. Therefore, according to the present embodiment, the carbon dioxide gas (CO2) that is separately supplied is used for cleansing the first solid-liquid separating film 25 and used as the pH adjuster 21, and thus it is possible to prevent inhibition of the denitrification reaction. At the same time, it is possible to reduce the amount of solidified form to be generated that is obtained at the time of incinerating and solidifying secondary waste produced when the reaeration-treated liquid 27 is treated. An apparatus for treating a radioactive nitrate waste liquid according to a fourth embodiment of the present invention is explained with reference to FIG. 4. The apparatus for treating a radioactive nitrate waste liquid according to the present embodiment is configured substantially identical to the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and thus like reference letters or numerals are denoted to like constituent elements of the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and redundant explanations thereof will be omitted. FIG. 4 is a schematic diagram of a configuration of the apparatus for treating a radioactive nitrate waste liquid according to the fourth embodiment of the present invention. As shown in FIG. 4, in an apparatus 10D for treating a radioactive nitrate waste liquid according to the present embodiment includes the gas circulating line 31B (that introduces the carbon dioxide gas (CO2)) only of the apparatus 10B for treating a radioactive nitrate waste liquid according to the second embodiment shown in FIG. 2, the pH sensor 36 that measures the pH of the denitrified liquid 23 in the denitrification tank 12C of the apparatus 10C for treating a radioactive nitrate waste liquid according to the third embodiment shown in FIG. 3, and the carbon-dioxide-gas supply-amount adjusting valve 37 that adjusts the amount of carbon dioxide gas (CO2) supplied to a denitrification tank 12D. When nitrogen gas (N2) and carbon dioxide gas (CO2) generated in the denitrification tank 12D are used and circulated into the denitrification tank 12D by using the gas circulating line 31B, it is possible to reduce the amount of carbon dioxide gas (CO2) introduced from outside. Further, the carbon dioxide gas (CO2) that is separately supplied is used for cleansing the first solid-liquid separating film 25 and used as the pH adjuster 21, and thus it is possible to prevent inhibition of the denitrification reaction. At the same time, it is possible to reduce the amount of solidified form to be generated that is obtained at the time of incinerating and solidifying secondary waste produced when the reaeration-treated liquid 27 is treated. This eliminates necessity of adding the pH adjuster such as sulfuric acid and hydrochloric acid. An apparatus for treating a radioactive nitrate waste liquid according to a fifth embodiment of the present invention is explained with reference to FIG. 5. The apparatus for treating a radioactive nitrate waste liquid according to the present embodiment is configured substantially identical to the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and thus like reference letters or numerals are denoted to like constituent elements of the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and redundant explanations thereof will be omitted. FIG. 5 is a schematic diagram of a configuration of an apparatus 10E for treating a radioactive nitrate waste liquid according to the fifth embodiment of the present invention. As shown in FIG. 5, the apparatus 10E for treating a radioactive nitrate waste liquid according to the present embodiment uses acetic acid as the carbon source 22, and includes the pH sensor 36 that measures the pH of the nitrate waste liquid 11 in the denitrification tank 12A of the apparatus 10A for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1 and a carbon-source supply-amount adjusting valve 39 that adjusts the amount of the carbon source 22 supplied to a denitrification tank 12E in an acetic-acid supplying line. The acetic acid used as the carbon source functions also as the pH adjuster, and thus it is possible to reduce the amount of or eliminate necessity of the pH adjuster such as sulfuric acid and hydrochloric acid. When the pH in the denitrification tank 12E measured by the pH sensor 36 is equal to or more than 9.0, the carbon-source supply-amount adjusting valve 39 is opened to lower the pH in the denitrification tank 12E, and when the pH in the denitrification tank 12E measured by the pH sensor 36 is equal to or less than 8.0, the carbon-source supply-amount adjusting valve 39 is closed to raise the pH in the denitrification tank 12E. In this manner, it is possible to prevent a decrease in pH caused due to excessive addition of acetic acid supplied as the carbon source and a decrease in performance inherent in nitrogen caused along therewith. Thus, when the pH in the denitrification tank 12E is measured by the pH sensor 36 to constantly control the pH in the denitrification tank 12E, the microorganism in the denitrification tank 12E can perform a denitrification reaction. Accordingly, it is possible to prevent inhibition of the denitrification reaction by the microorganism in the denitrification tank 12E. In the present embodiment, the acetic acid as the carbon source 22 is never added excessively, and thus it is possible to reduce the amount of acetic acid to be used. That is, in the conventional technique, when such an adjustment is not performed, the redundant acetic acid is dissolved in the reaeration tank 14 and a part thereof is left as the redundant sludge 26B. On the other hand, according to the apparatus 10E for treating a radioactive nitrate waste liquid of the present embodiment, the amount of acetic acid flown out to the reaeration tank 14 is reduced, and thus it is possible to reduce the amount of redundant sludge to be generated in the reaeration tank 14. An apparatus for treating a radioactive nitrate waste liquid according to a sixth embodiment of the present invention is explained with reference to FIG. 6. The apparatus for treating a radioactive nitrate waste liquid according to the present embodiment is configured substantially identical to the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and thus like reference letters or numerals are denoted to like constituent elements of the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and redundant explanations thereof will be omitted. FIG. 6 is a schematic diagram of a configuration of the apparatus for treating a radioactive nitrate waste liquid according to the sixth embodiment of the present invention. As shown in FIG. 6, in an apparatus 10F for treating a radioactive nitrate waste liquid according to the present embodiment, instead of nitrogen gas (N2), air is supplied to the denitrification tank 12E. Furthermore, the apparatus 10E includes a gas circulating line that circulates the nitrogen gas and the carbon dioxide gas produced in the denitrification tank 12F into the denitrification tank 12F. That is, the apparatus 10F for treating a radioactive nitrate waste liquid according to the present embodiment includes an air-supplying line 41 that supplies air instead of nitrogen gas to the denitrification tank of the apparatus 10A for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and the gas circulating line 31B that circulates the nitrogen gas (N2) and the carbon dioxide gas (CO2) produced in the denitrification tank of the apparatus 10B for treating a radioactive nitrate waste liquid according to the second embodiment shown in FIG. 2 into the denitrification tank 12F. Further, in the apparatus 10F for treating a radioactive nitrate waste liquid according to the present embodiment, an air-supply-amount adjusting valve 42 is arranged in the air-supplying line 41, and the air-supply-amount adjusting valve 42 is repeatedly opened and closed at regular intervals. The valve 42 can be opened and closed, for example, for about two to five minutes a week, to supply air intermittently to the denitrification tank 12F. For example, when acetic acid and glucose are used as the carbon source, if an oxidation reduction potential (ORP) decreases, the reaction progresses with the organic substance only and the nitrate is not used. Therefore, there is a case that the nitrate does not decrease; however, by supplying a slight amount of oxygen as mentioned above, the decrease in the denitrification performance can be prevented. In the present embodiment, acetic acid is used as the organic acid; however, the present invention is not limited thereto. For example, organic acid such as formic acid and propionic acid; and sweetener such as glucose, fructose, maltose, sucrose, and galactose can be also used. It is possible to increase the types of carbon sources that can be used to treat the high-concentration nitrate waste liquid 11. A reductive atmosphere in the denitrification tank 12F is mitigated, and thus generation of hydrogen sulfide (H2S) can be prevented and it is possible to suppress the corrosion of piping of a reaction container, for example. An apparatus for treating a radioactive nitrate waste liquid according to a seventh embodiment of the present invention is explained with reference to FIG. 7. The apparatus for treating a radioactive nitrate waste liquid according to the present embodiment is configured substantially identical to the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and thus like reference letters or numerals are denoted to like constituent elements of the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and redundant explanations thereof will be omitted. FIG. 7 is a schematic diagram of a configuration of the apparatus for treating a radioactive nitrate waste liquid according to the seventh embodiment of the present invention. As shown in FIG. 7, in an apparatus 10G for treating a radioactive nitrate waste liquid according to the present embodiment, instead of nitrogen gas, air is supplied to the denitrification tank, and includes the gas circulating line 31B that circulates nitrogen gas (N2) and carbon dioxide gas (CO2) produced in a denitrification tank 12G into the denitrification tank 12G. That is, the apparatus 10G for treating a radioactive nitrate waste liquid according to the present embodiment is configured by adding, in the denitrification tank 12G, an oxidation-reduction potential meter (ORP meter) 43 that measures the oxidation reduction potential (ORP) of the denitrified liquid 23, to the apparatus 10F for treating a radioactive nitrate waste liquid according to the sixth embodiment shown in FIG. 6. Based on a value of the oxidation-reduction potential measured by the oxidation-reduction potential meter 43 arranged in the denitrification tank, the air-supply-amount adjusting valve 42 is controlled so that the amount of air supplied to the denitrification tank 12G is adjusted. Specifically, it suffices that when the oxidation-reduction potential of the denitrified liquid 23 in the denitrification tank 12G that is measured by the oxidation-reduction potential meter 43 is −350 millivolts, preferably −300 millivolts, the air-supply-amount adjusting valve 42 is opened. On the other hand, it suffices that when the oxidation-reduction potential of the denitrified liquid 23 in the denitrification tank 12G that is measured by the oxidation-reduction potential meter 43 is −50 millivolts, preferably −100 millivolts, the air-supply-amount adjusting valve 42 is closed. When the oxidation-reduction potential of the denitrified liquid 23 in the denitrification tank 12G is measured by the oxidation-reduction potential meter 43, the oxidation-reduction potential of the denitrified liquid 23 in the denitrification tank 12G can be controlled within a constant range, and thus, even when the oxidation-reduction potential rises abnormally, a decrease in denitrification performance can be prevented. An apparatus for treating a radioactive nitrate waste liquid according to an eighth embodiment of the present invention is explained with reference to FIG. 8. The apparatus for treating a radioactive nitrate waste liquid according to the present embodiment is configured substantially identical to the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and thus like reference letters or numerals are denoted to like constituent elements of the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and redundant explanations thereof will be omitted. FIG. 8 is a schematic diagram of a configuration of the apparatus for treating a radioactive nitrate waste liquid according to the eighth embodiment of the present invention. As shown in FIG. 8, in an apparatus 10H for treating a radioactive nitrate waste liquid according to the present embodiment, an adjusting tank 51 that supplies a nitrate waste liquid (undiluted: salt concentration of 30 to 40%) 50 is arranged at an upstream of the denitrification tank 12A of the apparatus 10A for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1. The adjusting tank 51 includes an electrical conductivity meter (EC meter) 52 that measures a level of electric conductivity, an industrial-water introducing line 53, and an industrial-water supply-amount adjusting valve 54 which is arranged in the industrial-water introducing line 53 and which adjusts a flow rate of water to be diluted based on a measurement value of the electrical conductivity meter (EC meter) 52. The industrial-water supply-amount adjusting valve 54 arranged in the industrial-water introducing line 53 is controlled based on the EC value measured by the electrical conductivity meter (EC meter) 52 arranged in the adjusting tank 51 to adjust the flow rate of water supplied to the adjusting tank 51. Specifically, when the EC value measured by the electrical conductivity meter (EC meter) 52 is high, the industrial-water supply-amount adjusting valve 54 is opened, and when the EC value measured by the electrical conductivity meter (EC meter) 52 is low, the industrial-water supply-amount adjusting valve 54 is closed. In the apparatus 10H for treating a radioactive nitrate waste liquid according to the present embodiment, when a biological treatment is performed in a high salt concentration state of about 4%, for example, by diluting the nitrate waste liquid (undiluted: salt concentration of 30 to 40%), a control range of an EC value measured by the electrical conductivity meter (EC meter) 52 at this time is preferably from 50 to 70 mS/cm, and more preferably from 53 to 68 mS/cm. In the apparatus 10H for treating a radioactive nitrate waste liquid according to the present embodiment, a nitrate-waste-liquid supplying pump P3 is arranged in a nitrate-waste-liquid supplying line 55 that feeds the nitrate waste liquid 50 (undiluted liquid) to the adjusting tank 51. Based on the EC value measured by the electrical conductivity meter (EC meter) 52, the nitrate-waste-liquid supplying pump P3 is controlled, and a supply amount of the nitrate waste liquid 50 to be fed to the adjusting tank 51 is thus adjusted. Specifically, when the EC value measured by the electrical conductivity meter (EC meter) 52 is high, the nitrate-waste-liquid supplying pump P3 is stopped to halt supplying of the undiluted nitrate waste liquid (undiluted liquid) 50 to the adjusting tank 51, and when the EC value measured by the electrical conductivity meter (EC meter) 52 is low, the nitrate-waste-liquid supplying pump P3 is operated and the nitrate waste liquid 50 (undiluted liquid) is supplied to the adjusting tank 51. Accordingly, the concentration of a high-concentration nitrate waste liquid from treatment facilities is not always constant, and thus, even when the concentration of the nitrate waste liquid (undiluted liquid) 50 is not constant, it is possible to prevent transferring variation of the salt concentration in the nitrate waste liquid (undiluted liquid) 50 to the denitrification tank 12A. This makes it possible to prevent a significant decrease in denitrification performance caused when the salt concentration is greatly varied. Further, the nitrate waste liquid (diluted liquid) 11 with a constant salt concentration can be supplied to the denitrification tank 12A via the nitrate-waste-liquid supplying pump P4, and thus the denitrification performance can be stabilized. An apparatus for treating a radioactive nitrate waste liquid according to a ninth embodiment of the present invention is explained with reference to FIG. 9. An apparatus 10I for treating a radioactive nitrate waste liquid according to the present embodiment is configured substantially identical to the apparatus 10A for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and thus like reference letters or numerals are denoted to like constituent elements of the apparatus for treating a radioactive nitrate waste liquid according to the first embodiment shown in FIG. 1, and redundant explanations thereof will be omitted. FIG. 9 is a schematic diagram of a configuration of the apparatus for treating a radioactive nitrate waste liquid according to the ninth embodiment of the present invention. As shown in FIG. 9, in the apparatus 10I for treating a radioactive nitrate waste liquid according to the present embodiment, a denitrification tank 12I includes, instead of the first solid-liquid separating film 25, a weir 80 that prevents flowing-out sludge from leaking out. A stirring unit 81 that stirs the content inside of the denitrification tank 12I is also provided. Due to the installation of the weir 80, a part of the sludge in the denitrification tank 12I deposits and remains in the denitrification tank 12I. As a result, the sludge concentration can be increased, and thus the range over which the nitrate concentration can be applied is widened. A biological treatment system using an apparatus for treating a radioactive nitrate waste liquid according to a tenth embodiment of the present invention is explained with reference to FIG. 10. As the apparatus for treating a radioactive nitrate waste liquid according to the present embodiment, any one of the apparatuses for treating a radioactive nitrate waste liquid according to the first to ninth embodiments shown in FIGS. 1 to 9 can be used, and thus explanations thereof will be omitted. As shown in FIG. 10, a biological treatment system 60 according to the present embodiment includes a diluter 62 that dilutes the nitrate waste liquid (undiluted liquid) 50 with industrial water 61, a biological treatment device (any one of the apparatuses for treating a radioactive nitrate waste liquid according to the first to ninth embodiments) 63 in which nitrate ions in the diluted nitrate waste liquid 11 are microbially treated for reduction with nitrogen and reaeration, a sludge dewatering device 67 that separates sludge 65 and a treated liquid 70 from the biologically treated liquid treated in the biological treatment device 63 and dewaters the separated sludge 65, and an incinerator 69 that incinerates dewatered sludge 68. The diluter 62 operates in the adjusting tank 51 that dilutes an undiluted liquid as shown in FIG. 8 corresponding to the eighth embodiment described above. The biological treatment device 63 is an apparatus for treating a radioactive nitrate waste liquid including one of the denitrification tanks 12A to 12I and the reaeration tank 14. The sludge dewatering device 67 dewaters the redundant sludge discharged from the denitrification tank and the reaeration tank of the biological treatment device 63. The incinerator 69 incinerates the dewatered sludge 68 or secondary waste discharged from the sludge dewatering device 67 into incinerated ashes 71. In the biological treatment device 63 that biologically treats the nitrate waste liquid having a high salt concentration, at the time of dewatering the redundant sludge, the salt concentration in the water that adheres to the sludge is high, and thus an amount of secondary waste (incinerated ashes) occasionally increases. For this, in the sludge dewatering device 67, the industrial water 61 is added to flush the adhered liquid having a high salt concentration, and flushed cleansing water 72 is used as diluting water in the diluter 62 that dilutes the nitrate waste liquid 11. Therefore, a part of the industrial water 61 used in the diluter 62 is used for cleansing at the time of dewatering, and thus there is no need to use additional industrial water 61. Further, by decreasing the salt concentration in the adhering water, the amount of the secondary waste (incinerated ashes) can be decreased to about ½. As described above, by using the apparatus for treating a radioactive nitrate waste liquid according to the present invention, a waste liquid with a high nitrate concentration discharged from nuclear facilities such as a reprocessing plant can be microbially treated efficiently, and minute sludge can be prevented from flowing out with a treated liquid.
	abstract	A system with function of bending and elongation is used for discharging foreign matters from nuclear reactor vessel. The system includes an operating rod which includes a suction pipe, a bendable rod section connected to the suction pipe, and an expandable rod section connected to the bendable rod section; and a drainage pipe. A suction opening is disposed at the suction pipe and an electric valve is disposed at a connection of the suction opening and the suction pipe. A filter mesh is disposed in the suction pipe; a suction pump is disposed in the suction pipe; a touch switch is disposed on the filter mesh. A water inlet of the suction pump is connected to the suction opening, a water outlet of the suction pump is connected to the outside space of the suction pipe though the drainage pipe, and the electric valve is controlled by the touch switch.
	summary
050680820	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS In advance of the detailed description of the preferred embodiments of this invention, the principal theory or basic principle according to this invention will be described hereunder with reference to the accompanying drawings. Referring to FIG. 1A, it is assumed that a water gap having a width w exists between two fuel existing zones I and II represented so as to have a rectangular cross section, respectively, and width wf of the fuel existing zone I(II) is considerably larger than the width w of the water gap as viewed from the same direction. In this case, a relationship between the width w of the water gap and the variation of the multiplication factor of the neutron is represented by a graph shown in FIG. 1B, and a framed portion C in FIG. 1B is shown in FIG. 1C on an enlarged scale. Referring to FIG. 1B or 1C, the variation of the multiplication factor is shown by a dotted line representing a high temperature condition or shown by a solid line representing a cooled temperature condition, both lines starting from "0" reference point showing the condition of no water gap. In a case where the water gap exists in the axial and normal (usually, horizontal with respect to a light water reactor) directions in the fuel assembly, it is considerably difficult to design the fuel assembly so as to have a wide water gap zone, and in other words, the formation of a wide water gap in the limited area makes narrow the fuel existing zone, which results in the reduction of the heat generating zone. In one embodiment of this invention, wide and narrow areas of more than two kinds are interposed between or among the fuel rods, and accordingly, it may be necessary to clearly describe herein the characteristics of the area having a narrow width and being interposed between the fuel rods, these or this zone being called "interposed area" hereinafter. For this purpose, FIG. 1C is shown on an enlarged scale of the framed portion C of FIG. 1B, and theoretically calculating values of the variations of the multiplication factors in the case where the water gap has a width of 2 cm at the most show substantially identical curves to those shown in FIG. 1C. Namely, it will be understood from the curves in FIG. 1C that the variation of the multiplication factor in the high temperature operation cycle (i.e. void generating cycle) increases in the positive direction (i.e. the effective multiplication factor K.sub.eff increases) in accordance with the increasing in the width of the water gap, and in the cooled temperature condition, the effective multiplication factor K.sub.eff having been remarkably reduced when the width of the water gap exceeds about 1 cm in accordance with the increasing in the water gap width, thus resulting in the increasing in the subcriticality at the shut-down period of the reactor. The foregoing description was made on the view point of the relative operational variations of neutrons in two fuel zones with the water gap interposed, but the above described operation or function may be described on the basis of a mode in which the infinite multiplication factor K.sub.oo of the fuel assembly is divided into old known four factors. According to this mode, the curves shown in FIG. 1C will be represented by the variation of the characteristics regarding the thermal neutron utilization factor and the resonance escape probability. In the case where it is required to enlarge the width of the water gap in the interior of the fuel assembly without reducing the number of the fuel rods, it is obliged to reduce the widths of the gaps between the respective fuel rods, and the reduction of the gap widths therebetween results in the increase in the shielding effect of absorption. neutrons between the respective fuel rods in the resonance absorption As a result, the resonance escape probability is increased, and on the other hand, the thermal neutron flux ratio of the water gap zone with respect to the fuel zone is reduced and hence an effect for lowering the thermal neutron utilization factor is attained. Accordingly, it will be said that the curves of the graph shown in FIG. 1C are substantially determined by the offset effect of the water density dependency and the water gap width dependency of the above two effects. In order to ensure the gaps between the respective fuel rods and elongate the water gap width, it has to be done to remove the fuel rods (fuel material), and in such case, the variation of the resonance escape probability described above is not based on the shielding effect of the resonance neutrons and that probability is increased on the basis of the increase in the slow-down effect of the neutrons. Namely, in the case where the reactor is operated at a high temperature and the voids are generated, the moderator is in short supply, which is relieved by the introduction of the water gap, and as a result, the resonance escape probability is increased. The thermal neutron utilization factor will be substantially identical to that mentioned with respect to the above case. FIG. 2 is a schematic view of one embodiment of this invention made on the basis of the principal theory described hereinbefore, and FIG. 2A is an elevational section taken along the line A-A shown in FIG. 2B showing a plan view of a fuel assembly. Referring to FIG. 2, a fuel assembly 20 includes a water rod 23 in square cross section arranged in the central portion of the fuel assembly 20 and fuel rods 21 and 23 arranged, except the central portion, regularly in nine rows and nine lines as shown in FIG. 2B. These water rod 23 and fuel rods 21 and 22 are surrounded by a channel box 24 having upper and lower ends secured by upper and lower tie plates 25 and 26, respectively. The fuel rods 22 each in which an interposed member 27 is inserted are denoted by the letter P. The length of the interposed member 27 is suitably determined to be less than about 1/3 of the entire effective length H of the fuel rod, the length being about 30 to 60 cm in the illustrated embodiment. The interposed member 27 is so located in the .fuel rod 22 that the central portion of the interposed member 27 is positioned at a vertical portion having a height of about 3/4H as measured from the lower end of the effective length of the fuel rod. This is based on the fact that in a zone existing near the 3/4H portion and having a vertical width of about 1/4H the subcriticality is made small. Although the details of the interposed members 27 will be described hereinafter, as typical examples, the insertion of hollow tubes, insertion of graphites, insertion of ZrH.sub.2 pellets and the introduction of water will be listed up. Output power spike suppressing members 27a for preventing the output power spikes are inserted in the fuel rods 22 adjacent to the interposed members 27, respectively, which will be also described hereinafter. The fuel rods P into which the interposed members are arranged in cruciform (in the state of mutually perpendicular straight lines) with the water rod 23 located at the central portion thereof. Each group includes three fuel rods P and the neutron interaction between the respective groups is small in this embodiment, so that there is no need for locating the fuel rods P on the same level. The fuel rod P provided with an inserted or charged interposed member or material is hereinafter merely called "fuel rod P with the interposed member". FIG. 3A is a schematic view showing an example in which the fuel assemblies 20 each referred to hereinabove are applied to a water boiling reactor and FIG. 3B is a graph showing the void fraction in the axial direction of the reactor core and the subcriticality distribution. The interposed members 27 are arranged in portions hatched in FIG. 3A, and it is effective to accord with the location levels of the interposed members 27 in the respective fuel rods P, but it is not always necessary to accord with the locations thereof in the respective fuel assemblies 20. This is based on the fact that the width of the interposed area substantially corresponds at the most to the width of the water gap at the outer periphery of the fuel assembly and is also based on the operation due to the binding effect in the interior of the fuel rod. With the interposed area having a narrow width, the binding effect is substantially masked by the water gap zone disposed at the outer periphery of the reactor core to less affect on the fuel assemblies arranged neighborhood. The axial length of each of the interposed members is varied in accordance with the number of the fuel rods into which the interposed members are charged and the mode of arrangement, but usually, the axial length thereof is determined to be within 15 to 90 cm. Below 15 cm, much effect is not expected, whereas above 90 cm, the effect is not relatively proportional to the reduction of the amount of the heat generating material (fuel) to be charged. Since the interposed members are arranged at portions in the fuel assemblies at which the subcriticality is made smallest, the embodiment of this invention can be effectively operated. FIG. 4 shows a plan view of the second embodiment of the fuel assembly according to this invention, and the like reference numerals are assigned to elements or portions corresponding to those shown in FIGS. 1 to 3. This numerical coincidence will be applicable to other embodiments which will be described hereinafter. Concerning the fuel assembly shown in FIG. 4, one of the fuel rods P in each group, including three fuel rods P, is substituted for the usual fuel rod 21 arranged in each corner portion of the fuel assembly because the corner portion attains the effect for improving the shutdown margin of the reactor. The fuel rods P with the interposed members arranged at the corner portions of the fuel assembly serve remarkably to increase the effective multiplication factor K.sub.eff at the high temperature operation period of the reactor for the reason that there exists sufficient water around the fuel rods of the corner portions, thus easily generating the output power. This arrangement effectively utilizes this function of the fuel assembly. With the fuel rods P of the fuel assembly of this arrangement, when a burnable poison (for example, gadlinium Gd) is inserted into each fuel rod P at substantially the axially central portion by the amount such that the effect of the burnable poison is vanished at the end of the operation cycle, the aforementioned object can be attained without generating the output power spike. FIG. 5 is a plan view of the third embodiment according to this invention, in which the fuel rods P in which the interposed members are inserted are diagonally arranged in cruciform with the water rod 23 disposed at the central portion thereof, and accordingly, the fuel assembly of this embodiment includes sixteen fuel rods P with the interposed members and sixty usual fuel rods 21. FIG. 6 is a plan view showing the fourth embodiment according to this invention, in which two water rods 28 each having a relatively small diameter are arranged as referred to with respect to the conventional fuel assembly, and accordingly, in this embodiment, the fuel assembly includes thirteen fuel rods P with the interposed members and forty-nine usual fuel rods 21. The fuel assembly of this embodiment does not include a water rod having a large diameter, so that the mutual function or action between the respective groups of the fuel rods P with the interposed members (three or four rods P in each group) in the cruciform arrangement is slightly larger than that attained by the first embodiment, and in this embodiment, it will be better to locate the respective interposed members at substantially the same axial levels. FIG. 7 is a plan view of the fifth embodiment according to this invention, which utilizes two water rods 28 each having a small diameter as described with reference to the embodiment shown in FIG. 6. According to this embodiment, the fuel rods P with the interposed members are diagonally arranged, and fourteen fuel rods P with the interposed members and forty-eight usual fuel rods 21 are arranged. This fifth embodiment attains substantially the same function as that attained by the fourth embodiment shown in FIG. 6. FIG. 8 is a plan view of the sixth embodiment according to this invention which also utilizes two water rods 28 each having a small diameter as described with reference to the embodiments shown in FIGS. 6 and 7. The fuel assembly of this embodiment includes fuel rods P with interposed members gathered at the central portion thereof and comprises fourteen fuel rods P with the interposed members and forty-eight usual fuel rods 21. The fuel assembly of this embodiment attains substantially the same functions as those attained by the embodiments shown in FIGS. 6 and 7. FIG. 9 is a plan view of the seventh embodiment according to this invention, which utilizes two water rods 28 each having a small diameter as described with reference to the embodiments shown in FIGS. 6 and 7. In this embodiment, the fuel rods P with the interposed members are arranged in two rows each parallel to one diagonal line of the fuel assembly. The fuel assembly of this embodiment includes eight fuel rods P with the interposed members and fifty-four usual fuel rods 21 and attains substantially the same functions as those attained by the embodiments shown in FIGS. 6 and 7. Since the number of the fuel rods P with the interposed members are smaller than that of the former embodiments, the length of each interposed member may be made larger or a plurality of interposed members may be inserted into one fuel rod P. Accordingly, this embodiment can provide the improved shut-down margin and be suitable for adjusting the axial output power distribution. In other words, water is in short supply at the upper portion of the reactor core during the high temperature operation period, but the water shortage may be alleviated by the existence of the interposed members inserted in the fuel rods P. FIG. 10 is a plan view of the eighth embodiment according to this invention, which is applicable to the mode utilizing two water rods 28 each having a small diameter. In the fuel assembly of this embodiment, the fuel rods P with the interposed members are arranged in a double-cruciform, and the fuel assembly comprises twenty-six fuel rods P with the interposed members and thirty-six usual fuel rods 21. The fuel assembly also includes four units of 3.times.3 (three lines and three rows) sub-bundles at the axial level of the interposed member insertion height, which are widely separated by the existence of the interposed members, so that very large shut-down margin can be obtained on that level of height. The degree of the shut-down margin can be adjusted by regulating the axial length of the interposed members, and in case of the length of the interposed member being about 15 to 30 cm (usual), the remarkable effect can be expected. The heights of the fuel rods P in respective cruciform arrangements may be changed with each other as occasion demands, and this arrangement will effectively change the output power distribution. FIG. 11 is a plan view of the ninth embodiment according to this invention, in which a water rod 29 circular in section having a larger diameter and a size corresponding to four fuel rods arranged in a bundle in the fuel assembly of the embodiment shown in FIG. 10 is arranged at the central portion of the fuel assembly, which therefore includes sixteen fuel rods P with the interposed members and forty-four usual fuel rods 21. The fuel assembly of this embodiment attains function and effect slightly less than those attained by the embodiment shown in FIG. 10. FIG. 12 is a plan view of the tenth embodiment according to this invention, which corresponds to the arrangement of the fuel rods shown in FIG. 7, but the central four rods are substituted with one water rod 31 having a square cross section. The fuel assembly of this embodiment includes twelve fuel rods P with the interposed members and forty-eight usual fuel rods 21, and attains function and effect slightly less than those attained by the embodiment shown in FIG. 10. FIG. 13 is a plan view showing the eleventh embodiment according to this invention, in which fuel rods, i.e., fuel cells are arranged in eleven lines and eleven rows and a water rod 32 having a large diameter and a size corresponding to nine fuel rods in a bundle is located at the central portion of the fuel assembly. The fuel assembly of this embodiment includes sixteen fuel rods P with the interposed members arranged in cruciform (i.e. in lines normal to each other) with the water rod arranged centrally of the cross shape and ninety-six usual fuel rods 21. FIG. 14 is a plan view of the twelfth embodiment according to this invention, which is a modification of the embodiment shown in FIG. 13 having the fuel cells arranged in eleven rows and eleven lines. This modified embodiment includes the thirty-three fuel rods P with the interposed members and seventy-six usual fuel rods 21 with a water rod 33 square in section having a size corresponding to nine fuel rods in a bundle arranged at the center of the fuel assembly. The fuel zones are divided into nine sub-zones in accordance with the heights of the inserted locations of the interposed members to effectively increase the shut-down margin and improve the effective multiplication factor K.sub.eff of the reactor during the high temperature reactor operation period. FIG. 15 is a plan view of the thirteenth embodiment according to this invention, in which a water rod 34 square in cross section having a size corresponding to five fuel rods in a bundle is arranged at the central portion of the fuel assembly in a manner inclined by 45.degree. with respect to the fuel bundle, and the fuel bundle is divided into four sub-bundles by arranging twelve fuel rods P with the interposed members in cruciform. The fuel bundle of this embodiment includes sixty-four usual fuel rods 21. Accordingly, since the distances between the respective sub-bundles at the heights of the inserted location of the interposed members of the fuel rods P is large this embodiment can attain the effects for increasing the effective multiplication factor (hot K.sub.eff) at the high temperature operation period and for remarkably reducing the effective multiplication factor (cold K.sub.eff) at the low temperature operation period (i.e. attaining large shut-down margin of a reactor). With this embodiment, a remarkable effect ca be attained even by using the fuel rods P with the interposed members each having relatively a short length. FIG. 16 is a plan view of the fourteenth embodiment according to this invention, in which a water rod 35 square in cross section having a size corresponding to five fuel rods in a bundle is arranged at the central portion of the fuel assembly in a manner inclined by 45.degree. with respect to the fuel bundle, which is divided into nine sub-bundles. The sub-bundles are separated from each other through relatively wide gaps. Twelve fuel rods P with the inserted members are centrally arranged in cruciform and four fuel rods P are disposed one by one at the respective corner portions of the fuel bundle. This embodiment also includes sixty usual fuel rods 21. This embodiment can also attain the effects for increasing the effective multiplication factor (hot K.sub.eff) at the high temperature operation period and for reducing the effective multiplication factor (cold K.sub.eff) at the low temperature operation period (i.e. attaining large shut-down margin), but these effects are slightly less than those attained by the embodiment shown in FIG. 15. FIG. 17 is a plan view of the fifteenth embodiment according to this invention, which is a modification of the embodiment shown in FIG. 16 and in which a water rod 36 square in cross section having a size larger than that of the embodiment shown in FIG. 16 is arranged at the center portion of the fuel bundle and four additional fuel rods are arranged one by one at portions opposing to the respective sides of the square water rod 36. The fuel assembly of this invention includes sixteen fuel rods P with the interposed members and sixty usual fuel rods 21. According to this arrangement, since the centrally arranged square water rod 36 has a large size, the shut-down margin larger than that of the former embodiment can be attained. FIG. 18 is a plan view of the sixteenth embodiment according to this invention, which is a modification of the embodiment shown in FIG. 15 and in which a round water rod 37 having a large diameter and a cruciform wide water gap are offset with respect to the fuel bundles. This embodiment is effectively applicable to a reactor core in which widths of the water gaps are arranged at the outer periphery of the fuel assembly (usually called BWR-D lattice type reactor core). In the reactor core of this type, it is desired that a central tie rod for control rods arranged in cross shape is located on a left upper side as viewed in FIG. 18. Namely, it is desired to locate many fuel rods on the side of the wide water gap. The fuel assembly of this embodiment includes fourteen fuel rods P with the interposed members and sixty-three usual fuel rods 21. FIG. 19 is a plan view of the seventeenth embodiment according to this invention, in which four sub-bundles are arranged. Between the respective sub-bundles is defined a cross shaped gap 39 as non-boiling moderating water area. The fuel rods P with the interposed members are disposed in a bundle at the central corner portions of the respective sub-bundles and the fuel bundle totally includes twelve fuel rods P with the interposed members and fifty-two usual fuel rods 21. This embodiment attains effects for increasing the effective multiplication factor at the high temperature operation period and reducing the effective multiplication factor at the low temperature operation period (i.e. large shutdown margin). FIG. 20 is a plan view of the eighteenth embodiment according to this invention, which is constructed by nine sub-bundles 41 each comprising nine fuel rods 21. The respective sub-bundles 41 are separated by gaps 42 each having relatively wide width. The sub-bundle centrally arranged in this fuel assembly all comprises nine fuel rods P with the interposed members, and accordingly, the fuel bundle of this embodiment includes nine fuel rods P with the interposed members and seventy-two usual fuel rods 21. According to the arrangement of this embodiment, the effects substantially identical to those attained by the former embodiment can be also attained. FIG. 21 represents the nineteenth embodiment according to this invention, in which FIG. 21A is an elevational section taken along the line A-A shown in FIG. 21B and FIGS. 21B and 21C are cross sectional views taken along the lines B-B and C-C shown in FIG. 21A. The fuel assembly of this embodiment includes a square water rod 35 arranged in the central portion thereof and long fuel rod 21 and short fuel rod 43 (denoted by letter P in figures hereinafter) are regularly arranged in nine row and nine lines except for the central portion in which the square water rod 35 is located. The outer periphery of the fuel rods 21 and 43 is surrounded by a channel box 24 and the upper and lower ends of these fuel rods are secured by means of the upper and lower tie plates 25 and 26, respectively. Each short fuel rod 43 contains a stack (fuel pellets) 44, and an upper plenum 46 is disposed above the top portion of the stack 44 through an output power spike suppressing member 45 and a lower plenum 47 is disposed at the bottom of the stack 44. With the embodiment of this type, the output power is liable to be increased locally at a portion within about 5 cm, particularly 1 cm, above the location of the top fuel pellet contained in the short fuel rod 43, so that, in order to prevent the occurrence of this phenomenon, the output power spike suppressing member 45 is disposed above the top of the stack 44. The output power spike suppressing member 45 is for example constructed by a depleted uranium pellet, natural uranium pellet or annular pellet having a length of 0.5 to 5 cm, usually 1 to 2 cm, with a burnable poison contained in the central portion thereof (i.e. Gd.sub.2 O.sub.3 --UO.sub.2, Gd.sub.2 O.sub.3 --ZrO.sub.2, Gd.sub.2 O.sub.3 --Al.sub.2 O.sub.3, HfO.sub.2 --Yb.sub.2 O.sub.3, HfO.sub.2 --Dy.sub.2 O.sub.3 and the like) or pellet of non-burnable material (i.e. ZrO.sub.2, ZrO.sub.2 --GdO.sub.3, Al.sub.2 O.sub.3, Al.sub.2 O.sub.3 --Gd.sub.2 O.sub.3, HfO.sub. 2 --Yb.sub.2 O.sub.3, HfO.sub.2 --Dy.sub.2 O.sub.3 and the like). A gentle power increasing may be caused for a long fuel rod 21 adjacent to a vanishing rod portion disposed at the upper portion of the short fuel rod. In such a case, usually, there may be substantially no need for specific procedure for this phenomenon, but with a fuel assembly in future in which the fuel enrichment will be increased, such power increasing will be treated by slightly lowering the enrichment or inserting an annular fuel pellet with Gd.sub.2 O.sub.3 contained in the central portion thereof on the basis of the known technique. FIG. 22 represents the twentieth embodiment according to this invention, in which FIG. 22A shows an elevational section taken along the line A-A shown in FIG. 22B and FIGS. 22B and 22C are cross sectional views taken along the lines B-B and C-C shown in FIG. 22A, respectively. As shown in FIG. 22B, this embodiment includes four short fuel rod groups each comprising three fuel rods P and arranged in cruciform with a water rod 48 having a large diameter arranged at the central portion of the cross-shaped short fuel rods P. The short fuel rods P are also arranged at the respective corners of the square fuel bundle, but in this embodiment these fuel rods are not necessarily made to be short fuel rods P. However, there is such a tendency as that the coolant concentrates in the vanishing zone in a certain case, and in such a case, the cooling ability of the coolants at the corner portions may be made insufficient. Taking into consideration such matter, if the short fuel rods P are arranged at the respective corner portions of the fuel bundle, such problem will be effectively obviated. The fuel bundle of this embodiment includes sixteen short fuel rods P and sixty usual fuel rods 21. According to this embodiment, the gap width of the vanishing zone is made wide in comparison with the nineteenth embodiment, so that the increased effects can be attained even with the use of the same amount of the fuel. FIG. 23 is a plan view of the twenty-first embodiment according to this invention, which includes the short fuel rods P having four groups each comprising three short fuel rods P in cruciform with a square water rod 34 arranged centrally in the crossing short fuel rods P. This embodiment includes twelve short fuel rods P and sixty-four usual fuel rods 21. According to this embodiment, the size of the centrally arranged square water rod may be made further large, and since the concentrating characteristics of the coolant towards the neighborhood of the central water rod are small in comparison with the embodiment shown in FIG. 22, the short fuel rods P are not used for corner rods. FIG. 24 is a plan view of the fuel assembly of the twenty-second embodiment according to this invention, which is a modification of the embodiment shown in FIG. 21. In this embodiment shown in FIG. 24, two outer fuel rods of the respective groups of the short fuel rods P arranged in cruciform are constructed a the short fuel rods P and four corner fuel rods are substituted with the four short fuel rods P, respectively in comparison with the embodiment shown in FIG. 23. The fuel assembly of this embodiment thus includes twelve short fuel rods P and sixty-four usual fuel rods 21. FIG. 25 represents the twenty-third embodiment according to this invention, in which FIG. 25A shows an elevational section taken along the line A-A shown in FIG. 25B and FIGS. 25B, 25C and 25D are cross sectional views taken along the lines B-B, C-C and D-D respectively shown in FIG. 25A. This embodiment is a modification of the embodiment shown in FIG. 24, in which the centrally arranged water rod 49 is constructed by portions having diameters different in the axial direction, and for the provision of the different diameter portions, two types of short fuel rods P different in their axial lengths are used. Namely, in this embodiment, as shown in FIG. 25B, the fuel assembly includes four groups of the short fuel rods 43 (denoted as P1 in the illustration) each comprising three fuel rods P1 in cruciform with a water rod 49 having a large diameter arranged at the central portion of the cruciform short fuel rods P1 to thereby divide the interior of the fuel assembly into four sub-bundles. Gaps are defined between the respective sub-bundles and the groups of the short fuel rods P1. Four short fuel rods 51 (denoted as P2 in the illustrated embodiment) each having a length further shorter than that of the short fuel rod P1 are located at the inner corner portions of the respective sub-bundles facing the peripheral surface of the central water rod 49. An upper plenum 52 and a lower plenum 53 are also arranged at the upper and lower portions of each further short fuel rod P2, and an output power spike suppressing member 54 is inserted into the top portion of the pellets contained in the further short fuel rod P2. Accordingly, with the fuel assembly of this embodiment, the coolant has an occupation gradually increasing towards the upper portion thereof. This embodiment includes twelve short fuel rods P1, four further short fuel rods P2 and sixty usual fuel rods 21. FIG. 26 is a plan view of the twenty-fourth embodiment according to this embodiment, in which four groups of fuel rods each comprising three short fuel rods P1 and one fuel rod 21 are arranged in cruciform with a water rod 55 having a diameter slightly larger than that of the fuel rod arranged at the central portion of the crossing fuel rods, thus dividing the interior of the fuel bundle into four sub-bundles. This fuel bundle thus includes twelve short fuel rods P and sixty-eight usual fuel rods 21. FIG. 27 is a plan view of the twenty-fifth embodiment according to this invention, which is a modification of the embodiment shown in FIG. 21 and in which four fuel rods P each positioned at the central portion of each group of three short fuel rods P are substituted with the fuel rods positioned at the respective corner portions of the fuel bundle. This embodiment thus includes twelve short fuel rods P and sixty-four usual fuel rods 21. This embodiment may have a degradation in the improvement of the shut-down margin, but can attain the improved effective multiplication factor K.sub.eff at the power operation period because the output powers of the fuel rods placed between the short fuel rods P in the respective groups are relatively high. FIG. 28 is a plan view of the twenty-sixth embodiment according to this invention, in which short fuel rods P are diagonally arranged in cruciform with a water rod 35 square in cross section arranged at the center of the cruciform. The fuel assembly of this embodiment includes sixteen short fuel rods P and sixty usual fuel rods 21 and attains substantially the identical effect to that attained by the embodiment shown in FIG. 21. FIG. 29 is a plan view of the twenty-seventh embodiment according to this invention, in which two water rods 38 each of which has a small diameter, but not large diameter, as used in the conventional fuel assembly are arranged and the short fuel rods P are also arranged in cruciform. This fuel assembly includes thirteen short fuel rods P and forty-nine usual fuel rods 21. FIG. 30 is a plan view of the twenty-eighth embodiment according to this invention, in which, as shown in FIG. 29, two water rods 38 each having a small diameter are arranged and the short fuel rods P are also arranged diagonally. This embodiment includes fourteen short fuel rods P and forty-eight usual fuel rods 21. FIG. 31 is a plan view of the twenty-ninth embodiment according to this embodiment, in which two water rods 28, as represented by the embodiment shown in FIG. 11, each having a small diameter are arranged and two rows of short fuel rods P are arranged in parallel to one diagonal line. This embodiment includes eight short fuel rods P and fifty-four usual fuel rods 21. FIG. 32 is a plan view of the thirtieth embodiment according to this invention, in which two water rods 28 each having a small diameter are arranged and the short fuel rods P are arranged in double-cross shape. This embodiment includes twenty-six short fuel rods P and thirty-six usual fuel rods 21. The fuel assembly of this embodiment comprises, at the height of the vanishing zone, substantially four units of 3.times.3 (three rows and three lines) sub-bundles for attaining large shut-down margin. FIG. 33 is a plan view of the thirty-first embodiment according to this invention, in which a water rod 29 having a large diameter and a size corresponding to four fuel rods in a bundle centrally arranged in the embodiment shown in FIG. 32 and the short fuel rods P arranged on the side of the channel box in FIG. 32 are substituted with the fuel rods each having a normal axial length. This embodiment includes sixteen short fuel rods P and forty-four usual fuel rods 21. FIG. 34 is a plan view of the thirty-second embodiment according to this invention, which is obtained by applying the embodiment shown in FIG. 28 to 8.times.8 (eight lines and eight rows) fuel assembly, and in which a square water rod 56 having a size corresponding to four fuel rods in a bundle is arranged centrally in that embodiment. This fuel assembly includes twelve short fuel rods P and forty-eight usual fuel rods 21. FIG. 35 is a plan view of the thirty-third embodiment according to this invention, which includes fuel rods arranged in nine rows and nine lines and a water rod 48 having a large diameter and a size corresponding to nine fuel rods in a bundle and being arranged at the central portion of the fuel assembly. The short fuel rods P are also arranged in cruciform with the water rod 48 arranged centrally and the short fuel rods P are further located at (2,2) portions (intersecting portions of second row and second lines) from the outer sides of the fuel bundle. According to this arrangement, the pressure-loss resistance at the upper portions of the (2,2) positions of the short fuel rods P is reduced and further improved the shutdown margin. The fuel assembly of this embodiment includes sixteen short fuel rods P and ninety-six usual fuel rods 21. FIG. 36 is a plan view of the thirty-fourth embodiment according to this invention, which includes fuel rods arranged in nine rows and nine lines with a water rod 33 square in cross section having a size corresponding to nine fuel rods in a bundle and arranged at the central portion of the fuel assembly. The interior of this fuel bundle is divided into nine sub-bundles by arranging thirty-six short fuel rods P in a manner as shown. The bundle in the vanishing zone at the upper portion of the fuel assembly is divided into a plurality of small bundles so as to attain the further improved effects. FIG. 37 is a plan view of the thirty-fifth embodiment according to this invention, which is a type in combination of the embodiments illustrated in FIGS. 21 and 22. Namely, a water rod 35 square in cross section having a size corresponding to four fuel rods in a bundle is disposed at the central portion of the fuel assembly with an inclination of 45.degree. with respect to the fuel bundle and the fuel assembly is divided into nine sub-bundles separated respectively by gaps each having relatively wide width. Twelve short fuel rods P are arranged in cruciform at the central portion of the fuel bundle, and this embodiment also includes sixty-four usual fuel rods 21. The water rod of this embodiment may be made slightly smaller in size. FIG. 38 is a plan view of the thirty-sixth embodiment according to this invention, which is an improved modification of the embodiment shown in FIG. 37. Namely, as shown, a square water rod 36 has a size slightly larger than that of the embodiment shown in FIG. 37, and further one fuel rod is disposed at a portion facing the side surface of the water rod 36. The fuel rods located at the respective outside corners are substituted with the short fuel rods P, and accordingly, the fuel assembly of this embodiment includes sixteen short fuel rods P and sixty usual fuel rods 21. According to this embodiment, since the centrally arranged square water rod 36 has a size larger than that of the former embodiment, the shut-down margin is made larger than that of the former embodiment. FIG. 39 is a plan view of the thirty-seventh embodiment according to this invention, which is a modification of the embodiment shown in FIG. 29 and in which water gap defined between a water rod 37 having a large diameter and a sub-bundle is offset with respect to the fuel bundle. Accordingly, this embodiment is effectively applicable to a reactor core in which the widths of the water gaps at the outer periphery of the fuel assembly are different (usually called BWR=D lattice type reactor core). This fuel assembly is arranged in the reactor core so that the gaps on the left side and the upper side, as viewed, have wide widths, whereby the output power distribution is made flat and further improved effects can be attained. This embodiment includes fourteen short fuel rods P and sixty-three usual fuel rods 21. FIG. 40 is a plan view of the thirty-eighth embodiment according to this invention, which is of the type applicable to a conventional fuel assembly. Namely, the fuel assembly of this embodiment comprises four sub-bundles 38 which are separated from each other by a cruciform gap 39 which is defined as a non-boiling moderating water zone. The fuel assembly includes twelve short fuel rods P which are arranged in a bundle at the corner portions facing the central crossing portion of the gap 39 of the respective sub-bundles and fifty-two usual fuel rods 21. FIG. 41 is a plan view of the thirty-ninth embodiment according to this invention, which is of the type applicable to a conventional fuel assembly. Namely, the fuel assembly of this embodiment includes nine sub-bundles 41 each comprising nine fuel rods and separated from each other by gaps 42 each having a relatively wide width. One sub-bundle arranged at the central portion of the fuel assembly comprises nine short fuel rods P and the other sub-bundles totally comprise seventy-two usual fuel rods 21. FIG. 42 is a plan view of the fortieth embodiment according to this invention, which includes two water rods 57 (A, B) having the same sizes, each corresponding to four fuel rods in a bundle and five fuel rods 22 (denoted as P in the illustrated embodiments hereinafter) provided with interposed members inserted therein respectively and disposed between the water rods A and B. The usual fuel rods 21 are arranged in the interior of the fuel assembly in nine rows and nine lines except the central locations of the water rods A and B and the five fuel rods P with the interposed members. Accordingly, the fuel assembly of this embodiment includes two water rods 57, five fuel rods 22 with the interposed members and sixty-eight usual fuel rods 21, and as a whole, the fuel assembly of this embodiment has a symmetrical arrangement with respect to the diagonal line connecting the upper left corner and the lower right corner as viewed in FIG. 42. In addition, according to this embodiment, the interposed members are inserted into the respective fuel rods P at portions each including the neighborhood of a portion having the height of 3/4H (H being the effective axial length of the reactor core) at which the maintenance of the shut-down margin is made difficult. As shown by dotted lines, a wide zone including the interposed member existing area at this height (3/4H) portion is located, this zone ranging between two water rods A and B each having a large diameter, thus improving the shut-down margin. Fuel rods denoted by letters Q and q surrounding the water rods A and B are usual ones, but the fuel rods Q have surface areas facing the water rods A and B larger than the surface areas of the fuel rods q, so that the fuel rods Q much contribute to the reactivity of the water rods A and B, than the fuel rods q. In addition, the thermal neutron fluxes in the fuel rods Q and q at the portions corresponding to the portions at which the interposed material is inserted into the fuel rods P are made high and hence the reactivity attained by the fuel rods is increased. Accordingly, the thermal neutron fluxes of the fuel rods Q and the fuel rods P at portions except the portions of the height corresponding to the location of the interposed members is inserted into the fuel rods P. Therefore, at these portions, the effective multiplication factor K.sub.eff will be intentionally increased or decreased. With this embodiment shown in FIG. 42, the short fuel rods of the type described hereinbefore will be substituted for the fuel rods 22 with the interposed members, and in such substituted arrangement, vanishing rods may be disposed at portions at which the maintenance of the shut-down margin is made difficult. FIG. 43 is a plan view of the forty-first embodiment according to this invention, in which, in comparison with the arrangement of the former embodiment shown in FIG. 42, one water rod A is shifted rightwardly, as viewed, by one line and the other water rod B is shifted downwardly, as viewed, by one row, and two fuel rods P with the interposed members are disposed between these water rods A and B. The fuel assembly of this embodiment thus includes two water rods each having a large diameter, two fuel rods P with the interposed members, and seventy-one usual fuel rods 21. The small shut-down margin zone at the upper portion of the reactor core is made large as shown by the dotted line in comparison with the former embodiment shown in FIG. 42. This embodiment may be effectively applicable to the reactor core in which the water gap widths at the outer periphery of the fuel assembly are different (usually called BWR-D lattice type reactor core) FIG. 44 is a plan view of the forty-second embodiment according to this invention, in which two groups A and B of water rods 28 (each denoted by letter W in the illustrated embodiment) each having a small diameter are arranged in triangular shapes respectively, and five fuel rods P with the interposed members are arranged between the respective water rod groups A and B. The usual fuel rods 21 are regularly arranged in nine lines and nine rows except the portions at which the water rods and the fuel rods P with the interposed members are located. The fuel assembly of this embodiment thus includes two water rod groups containing six water rods W each having a small diameter, five fuel rods P with the interposed members and the usual fuel rods 21. One water rod having a triangular outer configuration may be utilized instead of one group of three annular water rods. With this embodiment, the interposed members inserted into the fuel rods P are positioned at the axial height of about 3/4H (H being substantial effective axial height of the reactor core) at which the maintenance of the shut-down margin is made difficult and the water rod zone extending in a direction having an inclination of 45.degree. is formed. The interposed member of one fuel rod P is designed so as to have an axial length of about 1/4H, so that the reduction of the fuel inventory due to the location of the fuel rods P with the interposed members will merely correspond to 1.25 number of fuel rods and hence as a whole assembly, to 7.25 number of fuel rods. The fuel inventory of this embodiment is made to be large in comparison with the usual case in which about nine fuel rods are charged for the fuel inventory. FIG. 45 is a plan view of the forty-third embodiment according to this invention, which is an improvement of the embodiment shown in FIG. 44 for intending to make the further large shut-down margin. The fuel assembly of this embodiment includes two groups A and B of water rods W each comprising three water rods 28 each having a small diameter and each having a triangular shape, three fuel rods 58 (denoted by letter P1 in the illustrated embodiment) disposed between the water rod groups A and B, and eight fuel rods 59 (denoted by letter P2) are arranged in cruciform on the outer side of the square zone constituted by the two water rod groups A and B and the three fuel rods P1. The usual fuel rods 21 are also arranged in nine lines and nine rows except the location of the water rod groups and the fuel rods P1 and P2. Thus, the fuel assembly of this invention includes six water rods W each having a small diameter, eleven fuel rods P (comprising three fuel rods P1 and eight fuel rods P2) and sixty-four usual fuel rods 21. The fuel rods P1 are provided with the interposed members each positioned at the portion having a height of about 3/4H of the reactor core at which the maintenance of the shut-down margin is made difficult. The fuel rods P1 may be substituted with the short fuel rods referred to hereinbefore above which the vanishing rods are arranged. The fuel rods P2 are of the character substantially the same as that of the fuel rods P1, but each has the effective fuel length longer than that of the latter P1. According to this embodiment, the fuel rods 58 and the water rods 28 function as a large water rod as shown by dotted line at the high temperature operation period of the reactor core, whereby the effective multiplication factor K.sub.eff is increased, and the reactivity at the upper portion Of the reactor core is improved, the distribution in the output lower axis direction also being improved. On the other hand, under the cold operation condition, the large water rod is of the excessive water condition, so that the subcriticality is improved and the fuel rods P2 serve to increase that effect. FIG. 46 is a plan view of the forty-fourth embodiment according to this invention, which is similar to the embodiment shown in FIG. 45, but is different in that four water rods 28 (W) each having a small diameter are arranged separately as water rod zones A, B, C and D, and five fuel rods 58(P1) are disposed between these four water rods W at the central portion of the fuel assembly. Accordingly, the non-fuel area, formed by the water rods and the fuel rods with interposed members, occupies the area corresponding to nine fuel rods at the upper portion of the reactor core. The fuel rods 59(P2) are arranged in cruciform for further improving the effects. The fuel assembly of this embodiment thus includes four water rods 28 each having a small diameter, thirteen fuel rods P (comprising five fuel rods P1 and eight fuel rods P2), and sixty-four usual fuel rods 21. FIG. 47 is a plan view of the forty-fifth embodiment according to this invention, which is generally constructed by nine sub-bundles each including 3.times.3 (three lines and three rows) fuel rods, and water gaps 61 each having a relatively wide width are defined between the respective sub-bundles. Two water rods 57 (denoted by letters A and B) each having a large diameter are disposed at the crossing portions of the water gaps 61 as shown in FIG. 47, and five fuel rods 22 (denoted by letter P) are disposed between the respective water rods A and B. As described above according to this embodiment, since the water rods A and B are disposed at the relatively large water gap zones, the water rods can be designed as relatively large ones comparatively of the reduced amount of the fuel inventory and the concentration of cooling water which is usually liable to be concentrated to these portions will be effectively suppressed by the location of the large water rods. Accordingly, the water rods A and B form at the upper portion of the reactor core a wide non-fuel area, referred to hereinbefore, through the fuel rods 22 with the interposed members. Flow control members (FC) 62 each having a small diameter are also provided in this fuel assembly at the other water gap crossing portions normal to those at which the water rods A and B are arranged to suppress the concentrated flow of the cooling water to these portions. The large water rods A and B also serve to control the flow of the cooling water. Each flow control members 62 usually have fine diameter holes so that the cooling water can flow in the arrowed directions. The fuel assembly thus includes two large water rods 57, five fuel rods P with the interposed members, sixty-eight usual fuel rods 21, and two FCs 62. FIG. 48 is a plan view of the forty-sixth embodiment according to this invention, which is constructed by nine sub-bundles (each 3.times.3 fuel rods) as referred to with respect to the former embodiment shown in FIG. 47. Water gaps 61 each having a slightly wide width are defined between the respective sub-bundles, but the large water rods A and B are arranged more closely than those shown in FIG. 47 and mutually connected by a water rod connecting member also commonly serving as flow control fin 63 at the several axial portions. The flow control fin 63 serves to force back the cooling water which is liable to be concentrated to the central portion of the fuel assembly. The upper portion of the reactor core above the portion enclosed by dotted lines constitutes no fuel area. The respective three fuel rods 22 each provided with an inserted member are arranged along the lengthwise directions of the flow control fin 63. The fuel assembly of this embodiment thus includes two large water rods 57, six fuel rods with the interposed members, and sixty-eight usual fuel rods 21. FIG. 49 is a plan view of the forty-seventh embodiment according to this invention, which is of the type different from that of FIG. 47 or FIG. 48, and in which the fuel assembly takes rough and fine lattice of 4-1-4 type fuel rod arrangement. Four water rods 28(W) portion of the central block enclosed by dotted lines, which constitute the water rod zones A, B, C and D. Five fuel rods 22 (P) with the interposed members are located between these water rods 28. The upper portion of the reactor core and the portion enclosed by dotted lines having water rods having a small diameter are arranged at corner is constituted as no fuel area, and the fuel assembly of this embodiment thus includes four water rods 28 each having a small diameter, five fuel rods 22 with the interposed members, and seventy-two usual fuel rods 21. FIG. 50 is a plan view of the forty-eighth embodiment according to this invention, in which the fuel assembly takes rough and fine lattice of 4-1-4 type fuel rod arrangement as described with reference to the former embodiment of FIG. 49. Two large water rods 57 (A and B) each having a size corresponding to four fuel rods in a bundle are located inside sub-bundles (each 4.times.4 fuel rods) on one diagonal line, and five fuel rods 22 (P) with the interposed members are arranged between these water rods A and B. Although this embodiment is somewhat similar in arrangement to the embodiment shown in FIG. 42, the gaps between the respective fuel rods P are made relatively wide, so that no fuel area at the upper portion, enclosed by dotted lines, of the reactor core is also made large. The fuel assembly of this embodiment includes two large water rods 57, five fuel rods 22 provided with the interposed members and sixty-eight usual fuel rods 21. FIG. 51 is a plan view of the forty-ninth embodiment according to this invention. The fuel assembly of this embodiment is of rough and fine lattice of 5-4 type fuel rod arrangement. Water rods 57 each having a large diameter are located at the inner corner portions of non-symmetrical lattice assemblies as shown in FIG. 50 and fuel rods 22 are located at the inner corner portions of symmetrical lattice assemblies. This fuel assembly thus includes two water rods 57 each having a large diameter, two fuel rods 22 with the interposed members and seventy-one usual fuel rods 21 and is effectively applicable to the BWR-D lattice type reactor core. FIG. 52 is a plan view of the fiftieth embodiment according to this invention. The fuel assembly of this embodiment is of rough and fine lattice of 4-2-3 type fuel rod arrangement. Although this embodiment is similar in arrangement to that shown in FIG. 51, flow control members (FC)62 are arranged such as shown in FIG. 47. The fuel assembly of this embodiment includes two water rods 57 each having a large diameter, two fuel rods 22 with the interposed members and seventy-one usual fuel rods 21 and is effectively applicable to the BWR-D lattice type reactor core. FIG. 53 is a plan view of the fifty-first embodiment according to this invention. The fuel assembly of this embodiment is of rough and fine lattice of 4-3-2 type fuel rod arrangement and similar to the embodiment shown in FIG. 52, but the water rods A and B have diameters slightly different from each other by taking into consideration the affect of the lattice shapes. The fuel assembly of this embodiment includes two water rods 57 having different large diameters, two fuel rods 22 with the interposed members, seventy-one usual fuel rods 21 and two FCs 63, and is effectively applicable to the BWR-D lattice type reactor core. FIG. 54 is a plan view of the fifty-second embodiment according to this invention. The fuel assembly of this embodiment is of 10.times.10 type (ten lines and ten rows of fuel rods) and includes rough and fine lattice of 4-2-4 type fuel rod arrangement. Accordingly, large water rods 57 (A and B) are adopted, but the other arrangement is substantially identical to that shown in FIG. 52. This fuel assembly includes two water rods 57 each having a large diameter, two fuel rods 22 with the interposed members, ninety usual fuel rods 21, and two FCs 63. FIG. 55 is a plan view of the fifty-third embodiment according to this invention. The fuel assembly of this embodiment is of 10.times.10 type and includes rough and fine lattice of (5.times.5).times.4 type fuel rod arrangement. Accordingly, the water rods 57 (A and B) each having a further large diameter is applicable in comparison with the water rods shown in FIG. 51, but the other structures are substantially identical to those of FIG. 51. The fuel assembly of this embodiment includes two water rods 57, two fuel rods 22 with the interposed members, and ninety-eight usual fuel rods 21. FIG. 56 is a plan view of the fifty-fourth embodiment according to this invention, in which four sub-bundles 60 are arranged and cruciform water gap 64 located between these sub-bundles 60 is constructed as non-boiling moderating water area. Three fuel rods 22 with the interposed members and one water rod 28 having a small diameter are arranged in a bundle at each corner portion of each sub-bundle facing the central crossing portion of the water gap 64. The fuel assembly of this embodiment thus includes four water rods 28, twelve fuel rods 22 with the interposed members, and eighty-four usual fuel rods 21, and according to this invention, no fuel area enclosed by dotted lines in FIG. 56 is formed at the upper portion of the reactor core. With the aforementioned forty-first to fifty-fourth embodiments, short fuel rods may be utilized in substitution for the fuel rods each provided with an interposed member. FIG. 57 is a plan view of the fifty-fifth embodiment according to this invention, in which FIG. 57A shows an elevational section taken along the line A-A shown in FIG. 57B and FIGS. 57B, 57C and 57D are cross sectional, i.e. plan view taken along lines B-B, C-C and D-D, respectively, shown in FIG. 57A. In the fuel assembly of this embodiment, a water rod 35 square in cross section is arranged at the central portion thereof, and fuel rods 22 each having a length longer than that of the short fuel rod and provided with an interposed member 66, and short fuel rods 65 (denoted by letter P in the illustrated embodiment) are regularly arranged in nine lines and nine rows except the central portion at which the water rod 35 is arranged. The outer periphery of these fuel rods is surrounded by a channel box 24 and the upper and lower ends thereof are secured by upper and lower tie plates 25 and 26, respectively. The short fuel rod 65 is provided with an upper plenum 46 above the top of the stack (pellets) 44 by way of an output power spike suppressing member 45, and a lower plenum 47 is disposed at the bottom of the stack 44. A portion above the short fuel rod 65 is occupied by the coolant and steam voids to form a vanishing area in which no fuel rod exists. Main portion of a gas plenum is located below the stack 44 and a short one is auxiliarily located above the stack. As shown in FIG. 57C, in the portions taken along the line C-C in FIG. 57A, all fuel rods contain the inserted interposed members 66 each having a length of about 2 to 8 cm, which is approximately equal to or more than the thermal neutron diffusion length at the cold operation period, whereas it is approximately equal to or less than that length at the generation of voids during the high temperature operation period. Output power spike suppressing members 67 are inserted in each of the fuel rods so as to nip the interposed member 66 from the upper and lower sides thereof. The characteristics required for the output power spike suppressing members 67 are basically the same as those required for the output power spike suppressing members 45 located above the top portion of the short fuel rods 65. The reactor core according to this embodiment has such a nature as that the reactor output power locally increases in a portion within 5 cm, particularly 1 cm, from the top of the pellet stack of the short fuel rod 65, so that in order to suppress this local output power increasing, the output power spike suppressing member 45 is inserted above the top of the pellet stack. The output spike suppressing members 67 are also axially inserted into the pellets between which the interposed member 66 is inserted. The constructions or compositions of the output power spike suppressing members are substantially identical to those referred to with respect to the nineteenth embodiment shown in FIG. 21. According to this embodiment, the output power gently increases in the long fuel rods 22 adjacent to the vanishing rods 68 located at the upper portions of the short fuel rods. In such a case, usually, there is no need for taking any specific procedure or counteraction, but with respect to a fuel assembly expected in future in which the fuel will be gradually enriched, many known techniques may be applicable such that the enrichment will be slightly lowered or an annular fuel pellet containing Gd.sub.2 O.sub.3 in the central portion thereof will be inserted Supposing a case in which no output power spike suppressing member exists, the output power spike of the upper and lower portions of the relatively short interposed member in the axial direction (vertical direction) of the C-C cross sectional portion (shown in FIG. 57C) will tend to be large in comparison with the maximum output power spike in a case where the relatively axially long interposed member is inserted in a direction normal to the axial direction (D-D cross sectional portion as shown in FIG. 57D) in the relatively upper portion of the fuel assembly. For this reason, the former may be arranged in a portion (upper portion of the reactor core) which sufficiently retains the cooling ability of the coolant for ensuring the soundness condition of the fuel in case of emergency. FIG. 58A is a schematic elevational view of a BWR-type core to which the fuel assembly according to this invention is applied and FIG. 58B is a graph representing the relationship between the void fraction in the axial direction of the reactor core shown in FIG. 58A and the distribution of the subcriticality. The hatched portions x in FIG. 58A represent the vanishing areas of the short fuel rods 65 and the hatched portions y represent areas into which the interposed members 66 are inserted. The heights of these vanishing areas x and the interposed member inserted areas y are usually aligned respectively, but this design matter is not always necessary because the vanishing areas have long length in the axial direction For example, in case it is desired to make gentle the axial output power distribution, the lower ends of these areas will be changed, for example, by providing a stepped portion in the bundle or a stepped portion between the respective bundles. On the other hand, it is desired to align the heights of the interposed member inserted areas y because these areas have vertically short lengths. FIG. 58B shows the comparison result between this invention and the conventional technique, and as is apparent from FIG. 58B, the subcriticality according to this invention is flat and large and the void fraction is slightly reduced at the upper portion of the reactor core. FIG. 59 shows schematic views of the fifty-sixth embodiment according to this invention, in which FIG. 59A shows an elevational section taken along the line A-A shown in FIG. 59B and FIGS. 59B, 59C, 59D and 59E are cross sectional views taken along the lines B-B, C-C, D-D and E-E respectively shown in FIG. 59A. In the fuel assembly of this invention, four groups, each comprising three fuel rods, of long fuel rods 69 (denoted by letter P in the illustrated embodiment), each provided with an interposed member, are arranged in cruciform with a water rod 37 having a large diameter arranged at the central portion of the crossing long fuel rods 69, and four short fuel rods 65 (denoted by leter PV) are located at the four crossed portions (2.2) of the second lines and second rows taken from the outer peripheral sides of the fuel bundle. Although the use of these short fuel rods 65 is not always necessary, the problem of high probability of insufficient cooling capacity at the portions near the bundle corners can be effectively overcome by using short fuel rods PV at the portions (2.2) as described above. The fuel assembly of this embodiment includes sixteen long and short fuel rods P and PV and sixty long fuel rods 22 each provided with an interposed member having the interposed member existing area smaller than that of the long fuel rod P. According to this embodiment, since the width of the gap between the axial interposed member existing areas is made wider than that of the former embodiment, the improved effects can be attained in comparison with the former embodiment by utilizing the same amounts of fuels. FIG. 60 is a plan view of the fifty-seventh embodiment according to this invention, which is to be applicable to a conventional type fuel assembly (in which two water rods 28 each having a small diameter, but not water rods each having a large diameter, are arranged), and fuel rods 69 with the interposed members are arranged in cruciform. The fuel assembly of this embodiment thus includes fourteen fuel rods P, comprising short fuel rods or long fuel rods 69 provided with the interposed members, and forty-nine fuel rods 22 with the interposed members having the interposed member existing area smaller than that of the fuel rods 69. FIG. 61 is a plan view of the fifty-eighth embodiment according to this invention, which is applicable to an embodiment similar to that shown in FIG. 60 in which each pair of adjacent water rods 28 have a small diameter. The fuel assembly of this embodiment includes fourteen fuel rods P with the interposed members arranged diagonally and forty-eight fuel rods 22 with the interposed members having the interposed member existing area smaller than that of the fuel rods P. FIG. 62 is a plan view of the fifty-ninth embodiment according to this invention, which is also applicable to an embodiment similar to that shown in FIG. 60 in which two water rods 28 each having a small diameter are arranged. The fuel assembly of this invention includes eight fuel rods P with the interposed members in two parallel rows in the diagonal direction by every four fuel rods P and fifty-four fuel rods 22 with the interposed members having the interposed member existing area smaller than that of the fuel rods P. FIG. 63 is a plan view of the sixtieth embodiment according to this invention, which is applicable to an embodiment in which two water rods 28 each having a small diameter are arranged. The fuel assembly of this embodiment includes twenty-six fuel rods P with interposed members located in double-cross shape arrangement and thirty-six fuel rods 22 with the interposed members having the interposed member existing area smaller than that of the fuel rods P. The fuel assembly is constructed in the axial interposed member existing area by four sub-bundles each comprising 3.times.3 (three rows and three lines) fuel rods, thus maintaining the large reactor shut-down margin. FIG. 64 is a plan view of the sixty-first embodiment according to this invention, in which a water rod 37 having a large diameter and the size corresponding to four fuel rods arranged at the central portion of the fuel assembly of the embodiment shown in FIG. 63, and in which the fuel rods P with the interposed members located on the sides of the channel box are substituted with the fuel rods 22 with the interposed members having the interposed member existing area smaller than that of the fuel rods P. The fuel assembly of this embodiment thus includes sixteen fuel rods P and the forty-four fuel rods 22. FIG. 65 is a plan view of the sixty-second embodiment according to this invention, in which a square water rod 31 having the size corresponding to four fuel rods is arranged at the central portion of the fuel assembly This fuel assembly includes twelve fuel rods P arranged in cruciform in the diagonal directions and forty-eight fuel rods 22 having the interposed member existing area smaller than that of the fuel rods P. FIG. 66 is a plan view of the sixty-third embodiment according to this invention, in which a large water rod 36 substantially square in cross section is arranged at the central portion of the fuel bundle in an inclined manner by 45.degree. with respect to the bundle surface. Further one fuel rod is arranged at a position facing each side surface of the water rod 36, and further one fuel rod P with the interposed member is located at each corner portion of the fuel bundle. The fuel assembly of this embodiment thus includes sixteen fuel rods P and sixty fuel rods 2 with the interposed members having the interposed member existing area smaller than that of the fuel rods P. FIG. 67 is a plan view of the sixty-fourth embodiment according to this invention, which is a modification of the embodiment shown in FIG. 60 and in which the water gaps between the water rod 37 having a large diameter and the sub-bundles are offset with respect to the fuel bundle, and accordingly, this embodiment is effectively applicable to a reactor core in which the water gaps at the outer peripheral portions of the fuel assembly have widths different from each other (called BWR-D lattice type reactor core). According to this embodiment, the flat power distribution can be effectively obtained by arranging the wide water gaps at the leftside and upperside portions, as viewed. This fuel assembly includes fourteen fuel rods P with the interposed members and sixty-three fuel rods 22 with the interposed members having the interposed member existing area smaller than that of the fuel rods P. FIG. 68 is a plan view of the sixty-fifth embodiment according to this invention, which is applicable to a conventional type fuel assembly. Namely, the fuel assembly of this embodiment is divided into four small channels 13 between which cross-shaped water gap 14 is located as non-boiled moderator water area. The fuel rods P with the interposed members are disposed in a bundle at the corner portions of the respective channels 13 facing the central portion of the crossed water gap in the fuel assembly. The fuel assembly of this embodiment thus includes twelve fuel rods P and fifty-two fuel rods 22 with the interposed members having the interposed member existing area smaller than that of the fuel rods P. FIG. 69 is a plan view of the sixty-sixth embodiment according to this invention, which is also applicable to an embodiment of a conventional type fuel assembly. Namely, this fuel assembly is constructed by nine sub-bundles 15 each comprising nine fuel rods 22 and slightly wide gaps 16 are located between the respective sub-bundles. The fuel assembly of this embodiment includes nine fuel rods P with the interposed members arranged in the central sub-bundle and seventy-two fuel rods 22 with the interposed members having the interposed member existing area smaller than that of the fuel rods P. FIGS. 70A to 70E are elevational sections of fuel rods of different types all according to this invention. A fuel rod shown in FIG. 70A has an area including no fuel material in a bundle, the area having a length of 2 to 90 cm into which a graphite 71 is charged. The interposed member charged into the vanishing area and the interior of the fuel assembly so as to separate them in a direction normal to the axial direction thereof has relatively long length of about 19 to 90 cm, and the interposed member inserted into the fuel rod so as to separate it in the axial direction has relatively short length of about 2 to 8 cm. These dimensional relationships will be applied to the other embodiments of the fuel rods described hereinafter. The graphite 71 has excellent high temperature characteristics and less absorption of the thermal neutrons as well as function as a moderator. Al.sub.2 O.sub.3, ZrO.sub.2 or the like as low density (i.e. porous) material has good heat-proof characteristics in spite of less moderating characteristics. Thus, such material having less neutron absorption characteristics may be utilized instead of the graphite. Hollow graphite, hollow Al.sub.2 O.sub.3, hollow ZrO.sub.2, hollow natural uranium and hollow depleted uranium may be also utilized instead of solid graphite, and in case such hollow material is utilized, the hollow portion will be utilized as a gas plenum. The most important one of the characteristics required for this area is that this area has the thermal neutron absorption rate, at the final stage of the reactor operation cycle, smaller than those of the fuel areas positioned on both axial sides of this fuel area. With the fuel materials adjacent to the charged graphite 71, power peaks (spikes) are generated in a range of about 2 cm (at the uppermost, 5 cm), which is not advantageous for the soundness of the fuel. Accordingly, respective two fuel pellets 72 each having a length of about 2 cm and bearing the burnable poisons 72a are arranged only at portions near the axis thereof. These pellets 72 do not bear the burnable poisons at their outer peripheral portions, so that the power output with relatively small variation can be obtained throughout the whole reactor operation cycle, and it is desired to design the fuel rod, so that the poison absorption characteristics will be vanished when the reactor operation cycle approaches the final stage and the output power of these areas will gently increase. The neutron interaction (binding effect) in the horizontal fuel area arranged adjacent to an interposed area (in which the concentration of the fissile nuclide is low) is reduced, and as a result, the subcriticality of the reactor during the shut-down period can be made further large. The difference between the fuel rod shown in FIG. 70A and that shown in FIG. 70B resides in that a tube 74 made of zircaloy having a small neutron cross section is inserted in the fuel rod shown in FIG. 70B instead of the graphite 71, and in this example, many modifications will be considered, for example: (1) When it is used as a gas plenum, the tube 74 is made as an unsealed tube. (2) In case ZrH.sub.2 (zirconium hydride) is fitted with high concentration, it is desired to seal the same in the tube, and a relatively small gap is provided in the tube for utilizing the gap as the gas plenum for H.sub.2 gas slightly discharged from the ZrH.sub.2. (3) Be and BeO which are poisonous are preferably charged in the tube, and since He gas is generated in the reaction between the Be and the neutrons, a small plenum for He gas will be disposed in the tube. Between the zircaloy tube 74 and the fuel pellet 73 are inserted or charged a thermal insulation pellet 75, Al.sub.2 O.sub.3, ZrO.sub.2, depleted uranium and the like for improving the soundness of the fuel. It is preferred that the thermal insulation pellet 75 has small thermal neutron absorption characteristics at the final stage of the reactor operation cycle, and accordingly, such as burnable poison bearing Al.sub.2 O.sub.3 --Gd.sub.2 O.sub.3 pellet or depleted uranium UO.sub.2 --Gd.sub.2 O.sub.3 pellet will be preferably utilized. With the axially adjacent fuel pellets of the zircaloy tube 74, it is desired to arrange the pellets 72 each containing the burnable poison so as to have a length of about 2 cm (at the uppermost, 5 cm) from the end portion thereof. Although, in the embodiment of FIG. 70B, the pellet 72 in which the Gd pellet having a fine diameter is inserted is illustrated, the Gd may be blended entirely into the pellet 72. This matter is applicable to the fuel rods shown in FIGS. 70A and 70C. The difference between the fuel rod shown in FIG. 70B and that shown in FIG. 70C resides in that water is introduced in the fuel rod shown in FIG. 70C. Namely, with the fuel rod of FIG. 70C, water holes 76 are formed in upper and lower portions of the clad 70 at which the zircaloy tube 74 is located, and an intermediate plug 77 and the thermal insulation pellet 75 are disposed at the upper portion and the lower portion of the respective water holes 76. The fuel pellets 72 containing the burnable poisons are disposed at the upper and lower portions of the intermediate plug 77 and the thermal insulation pellet 75, and the fuel pellets 73 are disposed at further upper and lower portions of the fuel pellets 72 in a bundle as shown in FIG. 70C. The difference between the fuel rod shown in FIG. 70D and that shown in FIG. 70A resides in that the fuel rod shown in FIG. 70D is provided with an interposed layer 78 wherein a burnable poison is added to the graphite (or Al.sub.2 O.sub.3, ZrO.sub.2, Al.sub.2 O.sub.3 --ZrO.sub.2). According to this embodiment, the burnable poison is not added to the fuel, the fuel rod is itself easily manufactured. The difference between the fuel rod shown in FIG. 70E and the fuel rod shown in FIG. 70A resides in that the fuel pellets axially adjacent to the interposed graphite (or Al.sub.2 O.sub.3, ZrO.sub.2, Al.sub.2 O.sub.3 --ZrO.sub.2) are specified so that the power spike can be hardly generated throughout the reactor operation cycle. Namely, with this embodiment, composite pellets 79, each in which is inserted a fine pellet formed by sintering a mixture oxide of depleted uranium oxide and gadolinium into an annular pellet utilizing natural uranium oxide, are arranged with an interposed member 71 therebetween. The thus prepared and arranged composite pellets 79 are enclosed in cassette shape with a thin metallic sleeve such as made of pure zirconium material. According to this embodiment, the fuel rod can be easily and precisely assembled. Boron may be used in substitution for the gadolinium. The depleted uranium has a small heat generation after the neutron toxicity of the Gd.sub.2 O.sub.3 at the core portion has been vanished by the sintering process, so that high temperature condition is not created, and in addition, the annular portion is filled with the natural uranium, so that high temperature condition is also not created at this portion. FIG. 71 shows one example of a control blade utilized in a reactor core in which the fuel assembly according to this invention is mounted, in which FIG. 71A is a graph representing characteristic curves required for the control blade, FIG. 71B shows the characteristic curves discreted on the basis of the characteristic curves shown in FIG. 71A, and FIG. 71C shows elevational sections of right half and left half of two examples of the control blades embodied on the basis of the graph shown in FIG. 71B. As is apparent from FIG. 3B and FIG. 58B, in a nuclear reactor in which the fuel assembly according to this invention is provided, the position having a small subcriticality is shifted from a portion near the upper portion of the reactor core, as in the conventional reactor core, to a portion near the axially central portion of the reactor core. For this reason, it is preferred that the reactivity of the control blade is made maximum at substantially the central portion of the control blade (corresponding to substantially the axially central portion of the reactor core) when the control blade is entirely charged. The distribution of the neutron irradiation amount is not widely varied in comparison with the conventional technique. Accordingly, it is desired, as shown in FIG. 71B, that, in consideration of the case of dividing the function of the control blade in the axial direction, the rear end portion of the control blade when charged into the reactor core has a reactivity substantially identical to that of the conventional control blade, at the central portion thereof, the priority is given to the reactivity, and at the top end thereof when charged into the reactor core, the priority is given to the long life and the reactivity is made substantially identical to that of the conventional control blade. Taking above matters into consideration, with the embodiment of the left half control blade shown in FIG. 71C, a control blade 80 is substantially equally divided into axial three portions. The rear end portion, when inserted into the reactor core, (having a length of about 1/4 to 1/3 of the total length of the control blade) is constructed as an ordinary zone 81 having a reactivity substantially identical to that of a conventional control blade and this zone 81 is filled with B.sub.4 C utilizing natural boron. The central portion (having a length approximately corresponding to a distance between a portion axially apart from the top end of the rear portion by a distance of 1/3 of the entire axial length of the control blade and a portion axially apart from the top end thereof by a distance of 2/3 of the entire length) is constructed as a high reactivity zone 82 which is filled with .sup.10 B.sub.4 C having a high reactivity. The front end portion (having a length of about 1/3 of the entire axial length of the control blade from the top end of the front portion) is constructed as a high neutron irradiation zone 83 utilizing an Hf plate as neutron absorption member. Air gape 84 and 85, which are filled with reactor water in the reactor are formed in the ordinary zone 81 and the high neutron irradiation zone 83 at the central portions thereof along the axis of the control blade. A control rod 80 shown as a right half of the embodiment shown in FIG. 71C is also axially divided substantially equally into three portions. The rear end portion, when inserted into the reactor core, is constructed as an ordinary zone 86 in which SUS tubes filled with B.sub.4 C powders are arranged side by side in a bundle. The central portion is constructed as a high reactivity zone 82 in which B.sub.4 C powders fill in the horizontal holes provided in SUS plates. The front end portion is constructed by inserting Hf plates having a trap-shape into the SUS plates. Accordingly, as described hereinabove, in a nuclear reactor utilizing the fuel assembly according to this invention in which the portion at which the subcriticality is made small is shifted from the axially upper portion of the reactor core to the central portion thereof in comparison with the conventional fuel assembly, the reactor operation can be most suitably controlled by axially dividing the control blade substantially equally into axial three portions thereby to vary the neutron absorption characteristics. FIG. 72 is an illustration of an in-pile instrumentation to be arranged inside the nuclear reactor constructed by the fuel assembly according to this invention. The in-pile instrumentation includes a neutron source range monitor (SRM) 91 and a local power range monitor (LPRM) 92. The neutron source range monitor 91 is located at an interposed area 93 of a fuel assembly 90 in which a large neutron flux peak exists for increasing the neutron counting rate at the low neutron flux level. The four local power range monitors 92 are located in a fuel effective area of the fuel assembly with substantially equal intervals in the axial direction thereof, and as shown in FIG. 72, the local power range monitors 92 arranged near the interposed area 93 are separated therefrom by distances of h.sub.1 and h.sub.2, being more than 10 cm, to obviate the occurrence of errors caused by the thermal neutron flux peak.
	summary
046438467	summary	BACKGROUND OF THE INVENTION The present invention relates to a process for the safe treatment of radioactive sodium which was formed by the contamination with radioactive substances or by the radioactivation of liquid sodium used as sodium coolant for fast breeder reactors or in a fused fuel-liquid sodium interaction test. Conventional methods for the treatment of sodium, regardless of whether it is radioactive or not, include (1) a method wherein sodium is heated to be reacted with water and oxygen in air by combustion, yielding Na.sub.2 O, Na.sub.2 O.sub.3, and NaOH; (2) a method wherein sodium is reacted with steam to form NaOH; (3) a method wherein sodium is reacted with an alcohol to form an alcoholate; and (4) a method wherein sodium is reacted with carbon dioxide gas or dry ice to form sodium carbonate. However, these methods have the following problems which must be solved in treating radioactive sodium. In method (1), fine smoke generated during combustion comprises radioactive sodium, the recovery or removal of which is very difficult. In method (2), the reaction must be conducted with great care. In method (3), large amounts of an alcohol is necessary because the solubility of sodium in the alcohol is small, and the waste alcohol must be disposed after the proper treatment. In method (4), large amounts of a gas must be treated and sodium must be strirred under heating to proceed the reaction because unreacted sodium is likely to remain in the central part of a reacting vessel. The radioactivity of liquid sodium which was used in a fast breeder reactor and has reached a radioactive equilibrium is several m Ci/kg. Na. In the treatment of radioactive sodium having such radioactivity, particular attention should be given to the following points: (a) The amount of the radioactive substances entrained with an exhaust gas to be released to external environments should be as small as possible. (b) The reaction should be stable, controlled easily and securely, without a fear of explosion. (c) The amount of wastes, e.g. waste liquor, discharged out of the facilities should be as small as possible. (d) The process should be a simple one not necessitating so many secondary materials. (e) The apparaus should be made compact. (f) The radioactive corrosion materials entrained with the radioactive sodium should be simultaneously treated. (g) From the viewpoint of the stability during temporary storage, the maintenance in the form of an alkaline or salt solution or of powder has a danger of leakage or scattering so that the radioactive substances should be maintained in a more stable form for example, as a vitrified body. SUMMARY OF THE INVENTION An object of the present invention is to overcome the disadvantages of the conventional methods described hereinbefore. A further object of the present invention is to provide an improved process for the treatment of radioactive sodium, which can satisfy all of the requirements described above. In the process of the treatment of radioactive sodium according to the present invention, radioactive sodium amalgam is first formed by mixing radioactive sodium with mercury. The resulting radioactive sodium amalgam is then reacted with water to form mercury and radioactive sodium hydroxide. The thus formed mercury is recycled into the step of forming radioactive sodium amalgam to be mixed with the radioactive sodium. The radioactive sodium hydroxide, on the other hand, is solidified in the presence of a solidifying material to be confined in a stable solidified body. In one embodiment of the present invention, the radioactive sodium hydroxide may be solidified after it is neutralized with an acid. The solidifying material may be selected from a vitrifiable substance, e.g. glass, a plastics, a cement or an asphalt.
	description	This application claims benefit of priority from Japanese application number JP 2009-52614 filed Mar. 5, 2009, the entire contents of which are incorporated by reference herein. 1. Field of the Invention The present invention relates to a radiation source container that accommodates a radiation source capsule in which a radiation source is sealed, and to a method of extending the sealing life of a radiation source capsule accommodated in the radiation source container. 2. Description of the Related Art Measurement equipment employing radiation to measure for example the thickness, level, density or water content of a measurement subject by measurements utilizing radiation is known. Also, regarding the various types of such devices utilizing radiation, safety standards in relation to sealing of the radiation source are laid down in Japanese Standard JIS Z4821 and safety standards regarding the radiation source container that accommodates the sealed radiation source are laid down in Japanese Standard JIS Z4614, respectively. Examples are to be found in: Technical Explanation of Measurement Equipment Utilizing Radiation [online], 2007, Japan Electric Measuring Instruments Manufacturers' Association Inc [retrieved 29 Jan. 2009] Internet <URL: http://www.jemima.or.jp/> (hereinbelow referred to as non-patent reference 1). A conventional radiation source container using γ rays as the radiation source in accordance with these standards will now be described with reference to FIG. 1A and FIG. 1B. FIG. 1A is a plan view of a conventional radiation source container 100 in a condition with the container lid 7b removed from the direction of the top; FIG. 1B is a cross-sectional view seen from the direction of the broken line arrows Xa-Xb of FIG. 1A. In FIG. 1A and FIG. 1B, the radiation source container 100 comprises a radiation source capsule 1 that emits γ rays and that is hermetically sealed with metal such as stainless steel, a radiation source holder 4 that accommodates this radiation source capsule 1, having an aperture B that emits γ rays emitted from an radiation port A in the middle of the top of the radiation source capsule 1 unidirectionally and whereby screening is effected such that γ rays cannot leak except from this aperture B, and on which holder this radiation source capsule 1 is fixed in such a manner that it is difficult to attach or detach, and a radiation source holder accommodating container 7 that accommodates the radiation source holder 4. This radiation source holder accommodating container 7 comprises a radiation source holder accommodating container lid 7b having an irradiation window C provided with a window plate 7b1 formed of a metal plate, such as aluminum, that is transparent to γ rays, and a radiation source holder accommodating container main unit 7a of this radiation source holder accommodating container lid 7b: γ rays are emitted from this irradiation window C. The radiation source holder accommodating container 7 further comprises, in its interior, an electrically operated shutter 6 that is operated by a control signal from outside, and a manually operated shutter 5 that is operated by hand so as to screen emitted γ rays from the irradiation window C provided between the radiation port A of the radiation source capsule 1 and the irradiation window C. In more detail, the radiation source holder 4 comprises a capsule fixing member 2a comprising a lid 2a1 and a seat 2a2 whereby the radiation source capsule 1 is fixed in the interior of this radiation source holder 4, a holder side plate 4a with a lead plate stuck thereon so as to prevent leakage of γ rays from outside the aperture B, and a holder base plate 4b. Also, when the radiation source container 100 is stored or transported, the manually operated shutter 5 is manually operated to operate a screening plate 5a made of a lead plate or tungsten plate, to effect screening so that there is no leakage of γ rays emitted from the aperture B. When the radiation source container 10 is incorporated in a measuring instrument employing radiation, the electrically operated shutter 6 controls whether or not other control devices constituting the measuring instrument employing radiation are subjected to irradiation or screened therefrom; in this electrically operated shutter 6, a screening plate 6d made of a lead plate or tungsten plate is mounted on a mounting base 6c; the mounting base 6c is mounted on a rotary shaft of a rotary solenoid 6b and is rotated by this rotary solenoid (coil) 6b; emission of γ rays from the aperture B is controlled by opening/closing the screening plate 6d; irradiation of outside the radiation source holder accommodating container 7 by γ rays from the irradiation window C is thereby controlled. The radiation source capsule 1 accommodated in the radiation source container 100 constructed in this way is usually sealed by welding in a stainless steel capsule; however, the manufacturers of the radiation source capsule 1 recommend that the period of use of this welded seal should be no more than about 15 years; if the radiation source container is to be used for more than this, replacement of the radiation source capsule 1 is recommended. Also, techniques have been disclosed for improving the performance of a level meter employing such γ rays. An example is to be found in Japanese Patent Number 3063488 (hereinbelow referred to as patent reference 1). However, in the case of 241Am (radiation source), the half-life is 432 years and even in the case of 137Cs (radiation source) the half-life is 30 years. Thus, from the point of view of the half-life of the radiation source, it would be possible to employ the radiation source container for longer than the recommended period of use of the welded seal. Thus, replacing the radiation source capsule within the recommended period of use of the welded seal presents the problems not only that the task of replacing each one of devices that are employed in large quantity is considerable, but also that of wastefully discarding radiation capsules. Consideration has therefore been given to employing the radiation sources for longer by improving the recommended period of use of the seal by re-doing the welded seal of the entire radiation capsule. However, such recycled capsules are subject to the problem that the amount of radiation is reduced so that the initial amount of radiation cannot be obtained, and also the problem that the guarantee of sealing is unreliable, so that a product of low reliability is obtained. The present invention was made in order to solve the above problems, its object being to provide a radiation source container and method of extending the sealing life of a radiation source capsule accommodated in this radiation source container whereby an extended life of the radiation source capsule can easily be achieved, without needing to replace the radiation source capsule at the recommended period of use in regard to the welded seal of the radiation source capsule. In order to achieve the above object, a radiation source container according to the present invention is constructed as follows. Specifically, a radiation source container comprises: a radiation source capsule; a radiation source holder having an aperture that emits unidirectionally radiation emitted from a radiation port of before-mentioned radiation source capsule, that provides screening such that no leakage of radiation takes place except from this aperture, and that fixes this radiation source capsule in a manner that is difficult to attach or detach; a radiation source holder accommodating container that accommodates before-mentioned radiation source holder and that has an irradiation window that transmits radiation emitted from before-mentioned aperture; an attenuation plate provided between the radiation port of before-mentioned radiation source capsule and before-mentioned irradiation window, that attenuates beforehand the amount of radiation emitted from before-mentioned irradiation window; a shutter provided between before-mentioned attenuation plate and before-mentioned irradiation window and that screens the radiation emitted from before-mentioned irradiation window; and a capsule cover whereby, when the recommended use period in respect of the sealing performance of before-mentioned radiation source capsule has expired, before-mentioned attenuating plate is removed and the radiation port of before-mentioned radiation source capsule is resealed with the same material properties and the same thickness as before-mentioned attenuating plate, covering this entire radiation source capsule. In addition, in order to achieve the above object, a method of extending the sealing life of a radiation source capsule of a radiation source container according to the present invention comprises the following steps. Specifically, a method of extending the sealing life of a radiation source capsule that seals a radiation source accommodated in a radiation source container comprises: a step of providing before-mentioned attenuating plate that produces attenuation beforehand so as to present a value corresponding to the amount of attenuation of the amount of radiation emitted from the irradiation window of before-mentioned radiation source container in the recommended period of sealing use of before-mentioned radiation capsule; and a step of when the recommended use period in respect of the sealing performance of before-mentioned radiation source capsule has expired, removing before-mentioned attenuating plate and resealing the radiation port of before-mentioned radiation source capsule with the same material properties and the same thickness as before-mentioned attenuating plate, covering this entire radiation source capsule. With the present invention, a radiation source container that accommodates a radiation source capsule and method of extending the sealing life of its radiation source capsule is obtained whereby an extended life of the radiation source capsule can easily be achieved, without needing to replace the radiation source capsule at the recommended period of use in regard to the welded seal of the radiation source capsule. An embodiment of the present invention is described below with reference to the drawings. A radiation source container 10 will now be described with reference to FIG. 2A, FIG. 2B and FIG. 2C to FIG. 4. Portions in the embodiment shown in FIG. 2A, FIG. 2B and FIG. 2C that are the same as in the conventional container shown in FIG. 1A and FIG. 1B are given the same reference symbols and further description thereof is dispensed with. The points of difference of the embodiment shown in FIG. 2A, FIG. 2B and FIG. 2C compared with the conventional construction shown in FIG. 1A and FIG. 1B are that, at the stage where the radiation source container 10 is initially used, an attenuation plate 3 that attenuates the radiation passing therethrough to a preset amount of radiation is provided in the aperture B and, at the stage where the prescribed recommended period of use in regard to the welded seal of the radiation capsule 1 has reached the time where replacement is recommended, the capsule fixing member 2a and this attenuating plate 3 are removed, and the radiation source capsule is accommodated in a capsule cover 2b which is resealed by welding, before being again accommodated in the radiation source holder 4. Next, the principles of the present invention will be explained with reference to FIG. 3A, FIG. 3B, FIG. 3C, FIG. 3D and FIG. 3E and FIG. 4. FIG. 3B, FIG. 3C, FIG. 3D and FIG. 3E show a model view of a radiation source container 10 and FIG. 4 shows an example of the attenuation characteristic with time of the amount of radiation emitted from the radiation source capsule 1 during this period. When the attenuation plate 3 of thickness th, prescribed beforehand, is provided between the radiation port of the radiation source capsule 1 and the irradiation window of the radiation source container, the amount of radiation I emitted from the radiation source container is expressed by the following expression (1)I=I0·e−μΔt1·e−λt (1) where I0 is the amount of radiation emitted when no attenuation plate 3 of FIG. 3A is present λ is the decay constant of the sealed radiation source t is the time Δth is the thickness of the attenuation plate, and μ is the radiation absorption coefficient of the attenuation plate. As shown in FIG. 4, when the half-life T of the sealed radiation source is sufficiently longer than the recommended period of use q for which sealing is maintained, the amount of attenuation of the initial amount of radiation I0 at the time-point where the recommended period of use q has elapsed can be neglected. However, when the half-life T of the sealed radiation source is not sufficiently longer than the recommended period of use q for which sealing is maintained, the amount B of attenuation at the time-point where the recommended period of use q has elapsed cannot be neglected. For example, if the amount of radiation that is required to be emitted from this sealed radiation source is Ia and an extended life of twice the recommended period of use q is to be guaranteed, it is necessary to satisfy:It2=I0·e−μΔth·e−λt2≧Ia (2) Also, when set to the amount of attenuation B of the radiation produced by an attenuating plate 3 of thickness Δth, B is:B=I0·e−μ·Δth (3) Consequently, at the time point t2 where a period 2q of (t2−t0) has elapsed, if the necessary amount of radiation is to be guaranteed, it is necessary to guarantee that the initial value of the amount of radiation I0 at the time-point t0 is:I0≧Ia+3B (4) Accordingly, if a radiation capsule 1 of radiation amount satisfying this expression (3) is provided, and the attenuating plate 3 is removed and resealing is effected by the capsule cover 2b of the same thickness Δt1 and the same material properties as the attenuating plate 3 at the time-point t1 of the recommended period of use q of this welded seal, the amount of radiation at the time-point t2 is It2=I0−2B: thus, if the recommended period of use q of the welded seal of the resealed capsule cover 2b is about the same as or better than before, the recommended period of use q of this radiation source capsule 1 can be extended by a factor of two. Since this capsule cover 2b is manufactured as an overlay, it is necessary to make the accommodation space of the radiation capsule 1 of the radiation source holder 4 larger by an amount corresponding to the number of times of resealing. In this way, by selecting beforehand the sealed radiation taking into account the necessary period for which the amount of radiation is required and sealing with a specification equivalent to the recommended period of use q of the welded seal, the time required for a fresh evaluation of the extended life of the welded seal can be dispensed with, making it possible to achieve reliable extension of the life of the radiation capsule 1 in an effective fashion. Hereinabove, the case where a single attenuating plate was provided was described. However, further, it would be possible, when extending the life of the sealed radiation source, to provide two such attenuating plates 3, and to arrange for one of these attenuating plates to be removed at the time-point where the recommended period of use q in regard to sealing performance is reached, resealing then being performed by a new capsule cover 2b-1; then, further, at the time-point where the recommended period of use q in regard to sealing performance is next reached, this second attenuating plate 3 is removed, and resealing is effected with a new capsule cover 2b-2. Next, the capsule cover 2b and attenuating plate 3 constituted in accordance with the principles of the present invention explained above will be described. The capsule cover 2b constituted by a top capsule cover 2b1 and bottom capsule cover 2b2 is of a construction in which these can be easily mutually fitted by insertion, so that, when the radiation source capsule 1 is again accommodated, the weld-sealing operation can easily be performed. Also, since the radiation source capsule 1 that is secured by welding sealing of stainless steel is of already-known material properties and thickness, its extended life in the environment in which it is to be used can be ascertained beforehand, as a database, for each type of measuring equipment using radiation. Consequently, life extension of high reliability can be achieved by weld-sealing with a capsule cover 2b of the same thickness and same material properties, and, in addition, by employing an attenuating plate 3 of the same material properties and thickness. As described above, with the present invention, extension of life of the radiation source capsule can easily be achieved by providing an attenuating plate determined beforehand in accordance with the amount of radiation to be emitted, and repeatedly resealing a capsule cover of the same material properties and thickness as this attenuating plate. The present invention is not restricted in any way to the embodiments described above and the attenuating plate and capsule cover, so long as they are provided between the irradiation window of the radiation source container and the radiation port of the sealed radiation source, and the same material properties and same thickness are selected, could be implemented with various suitable modifications of shape and material properties in accordance with the manufacturing specification of the sealed radiation source, within a scope not departing from the gist of the present invention.
	summary
052079791	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to a nuclear fuel assembly and to the core of a reactor and, more particularly, to a fuel assembly and to a reactor core of a boiling water reactor. 2. Description of the Prior Art The fuel assembly used in a boiling water reactor typically includes fuel rods, water rods and upper and lower tie plates for supporting the upper and lower end portions of the fuel rods and water rods. The water rods are disposed among the fuel rods. The fuel rods comprise fuel rods containing a burnable poison (such as gadolinia (Gd.sub.2 O.sub.3)) and fuel rods not containing it. In a boiling water reactor, because the void fraction (ratio of steam volume to total volume) varies over the core height, the power distribution in the axial direction of the core is low at an upper portion and high at a lower portion. Therefore, a burnable poison is used not only to restrict an excessive reaction rate but also to control the power distribution in the axial direction of the core. Examples of such use of a burnable poison are represented by the fuel assemblies shown in FIGS. 2A and 2B and FIGS. 3A and 3B of U.S. Pat. No. 4,587,090. The first of these includes fuel rods whose gadolinia concentration (4.5 wt %) in the axial direction is uniform and fuel rods which are divided into the upper and lower regions at the position of 11/24 from the lower end of their effective fuel length and whose gadolinia concentration is higher in the lower region (4.5 wt %) than in the upper region (3.5 wt %). The second of these fuel assemblies includes fuel rods of the latter, divided type and fuel rods which contain gadolinia (4.5 wt %) in the upper region and do not contain any gadolinia in the lower region. Furthermore, the fuel assembly shown in FIG. 1 of Japanese Laid-Open Patent Specification No. 149588/1988 includes two kinds of fuel rods containing burnable poison. These fuel rods comprise fuel rods having a low gadolinia concentration and fuel rods having a high gadolinia concentration. In each gadolinia-containing fuel rod, gadolinia is distributed uniformly in the axial direction. These fuel rods, including those which do not contain gadolinia, are disposed in a rectangular array. In other words, each fuel rod is positioned at the point of intersection of a rectangular grid. In such a structure, the fuel rods having a low gadolinia concentration and the fuel rods having a high gadolinia concentration are disposed adjacent to one another in the rows and columns of the array, which are parallel to the side surfaces of the lower tie plate (i.e. in an X direction or in a Y direction orthogonal to the X direction). Mention is also made of Japanese Laid-Open Patent Specification No. 217186/1987 referred to below. In the boiling water reactor having in its core the fuel assembly shown in FIG. 2A (using the fule rods shown in FIG. 2B) or the fuel assembly shown in its FIG. 3A (using the fuel rods shown in FIG. 3B) of U.S. Pat. No. 4,587,090, the axial power peaking is high at the initial stage of an operation cycle. Axial power peaking describes the peak value of power at one axial location of the core, i.e. deviation from a uniform power distribution. Particularly when the latter fuel assembly (FIGS. 3A and 3B), in which the number of gadolinia-containing fuel rods is smaller in the lower region than in the upper region, is employed, the axial power peaking is very high at the initial stage of the operation cycle. The operation of the core is conducted at a high average void fraction with axially skewed power distribution in the first half of the operation cycle, and the average void fraction becomes low since the power of the lower region, where burn-up has proceeded, becomes low at the final stage of the operation cycle. A spectral shift effect occurs by changing in this manner the average void fraction during the operation cycle. In the fuel assembly described in Japanese Laid-Open Patent Specification No. 149588/1988 with fuel rods having a low gadolinia concentration and fuel rods having a high gadolinia concentration, the gadolinia concentration of the former is below 60% of that of the latter. As a result of studies, the present inventors have found that in such a fuel assembly, the axial power peaking at the initial stage of the operation cycle (0 to about 2 GWd/st) reduces more than in the two fuel assemblies disclosed in U.S. Pat. No. 4,587,090. SUMMARY OF THE INVENTION However, in order further to improve characteristics such as a linear heat generation rate and thermal hydraulic stability, it is desired to reduce the axial power peaking at the initial stage of the operation cycle. It is an object of the present invention to provide a fuel assembly which can reduce the axial power peaking at the initial stage of the operation cycle. In the fuel assembly of the invention, the fuel rods containing the burnable poison include (a) at least one fuel rod which has the minimum burnable poison concentration inside the fuel assembly at a lower region and (b) other fuel rods. Fuel rods not containing the burnable poison are disposed between the fuel rods containing burnable poison. Since the fuel rods having the minimum burnable poison concentration in the lower region are disposed spaced-apart from the other fuel rods containing burnable poison, reactivity in the lowest region is efficiently reduced by the rods having burnable poison of the minimum concentration and thus the axial power peaking at the initial stage of the operation cycle can be reduced effectively.
	abstract	An apparatus for containing and cooling enriched water for the production of activated fluorine (18F). A target assembly includes internal cooling channels in which developed flow of a coolant removes the heat from the target liquid in the target chamber. In one embodiment, the target assembly is fabricated of tantalum.
	description	An electrostatic multi-lens according to the present invention which has high charged beam position stability and lens focusing characteristics, and can minimize crosstalk between lenses, and its application method will be explained below. (Basic Structure of Lens) FIG. 1 shows the lens electrode structure of a resistor electrode lens 1 as an electrostatic lens according to the present invention. This resistor electrode lens 1 is comprised of a lens electrode substrate 10 formed of two high-resistance electrode substrates 2 and 3, an upper electrode substrate 7, and a lower electrode substrate 8. Each electrode substrate is formed with lens apertures 9 at the positions of beam axes 11 where charged beams pass. On each of the high-resistance electrode substrates 2 and 3, the lens apertures 9 are formed on a substrate made up of an insulator 4, and their inner walls are covered by a high-resistance layer 5. The high-resistance layer 5 is electrically connected to a wiring electrode 6 formed on the substrate of the insulator 4. The high-resistance electrode substrates 2 and 3 are bonded to each other to form the lens electrode substrate 10. The upper and lower electrode substrates 7 and 8 each made up of a low-resistor are respectively bonded to the upper and lower surfaces of the substrate 10. FIG. 4A is a sectional view of one lens of the resistor electrode lens 1 shown in FIG. 1. The electric field intensity distribution in the lens aperture 9 is formed by the potential of the wiring electrode 6 given by a lens control power supply 41, and the potential of the high-resistance layer 5 generated between the upper electrode substrate 7 and the ground potential of the lower electrode substrate 8. As can be seen from FIG. 4A, since the beam axis 11 of a charged beam which passes through the lens is surrounded by the high- and low-resistors, even if a scattered beam strikes the side wall in the lens, the charge is relieved via the high-resistance layer 5, thus preventing charge-up in the lens. Hence, position instability of the charged beam and deteriorated lens focusing characteristics can be prevented. Since the high-resistance layer 5 in each lens aperture of the resistor electrode lens 1 has an effect of shielding the influence of the electric fields of neighboring electrodes, it can prevent the influence of crosstalk of neighboring lens electric fields of the multi-lens. Furthermore, as shown in FIG. 1, the upper and lower electrode substrates 7 and 8, each of which is made up of a low-resistor material, are provided to the upper and lower surfaces of the lens electrode substrate 10 so as to suppress the influence of crosstalk between neighboring lenses due to the electric field distribution which extends from each lens aperture 9 to outside the lens. If T represents the thickness of each of the upper and lower electrode substrates 7 and 8, and D represents the diameter of the lens aperture 9, crosstalk between neighboring lens apertures 9 can be prevented by setting Txe2x89xa70.3D. A feature of the lens shape will be explained below. The structure of the resistor electrode lens 1 has a simple cylindrical shape, and can assure a broad effective area of the lens aperture 9 since it has no complicated shield electrode unlike in the prior art shown in FIG. 7. This means that not only the electric field distribution uniformity in the lens aperture 9 is improved to minimize lens aberrations of the electrostatic lens but also the mechanical required precision of the electrode shape can be relaxed. FIG. 2 shows an example of the resistor electrode lens 1 having a higher lens layout density. In the conventional electrostatic lens, since the lens apertures cannot be formed at nearby positions due to the presence of shield electrodes, lenses are laid out at intersections between lines X1, X2, and X3, and Y1, Y2, and Y3 in FIG. 2. By contrast, since the resistor electrode lens 1 of the present invention has no shield electrode, electrostatic lenses can also be laid out at the intersections between lines X1xe2x80x2 and X2xe2x80x2, and Y1xe2x80x2 and Y2xe2x80x2. This layout forms a face-centered structure, and can improve the layout density to twice or more that using the conventional electrostatic lenses. (Optimization of On-axis Electric Field Distribution) Another feature of the resistor electrode lens 1 lies in that the on-axis electric field distribution can be arbitrarily determined by changing the resistance of the high-resistance layer 5 in each lens aperture 9 along the direction of the beam axis, thus forming an electric field distribution suitable for the focusing characteristics of the electrostatic lens. FIG. 5 shows the surface potential of the high-resistance layer 5 required for obtaining a given focal length when the resistance of the high-resistance layer 5 of the resistor electrode lens shown in FIG. 4A is changed. This example shows the characteristics when the diameter of the lens aperture 9 is 160 xcexcm, the thickness of each of the high-resistance electrode substrates 2 and 3 is 300 xcexcm, the substrate thickness of each of the upper and lower electrode substrates 7 and 8 is 100 xcexcm, and the electrostatic lens forms a deceleration system. Lines a to c in FIG. 5 represent the positions and surface potentials of the high-resistance layer 5 corresponding to three different resistor distributions required for obtaining a focal length=1 m of an electron beam of 50 KeV. Line a represents a case wherein the high-resistance layer 5 has a constant resistance along the beam axis. Lines b and c represent cases wherein the high-resistance layer 5 has two different resistances along the beam axis. Line b represents a case wherein the resistance on the entrance and exit sides of the beam is nearly twice that on the wiring layer side of the lens center, and line c represents the case the resistance on the wiring layer side is nearly xc2xd that on the entrance and exit sides of the beam. The results of the beam focusing characteristics are 0.16 xcexcm in case of line a, 0.14 xcexcm in case of line b, and 0.23 xcexcm in case of line c. In these cases, a lens electric field formed by the resistance distribution of line b exhibits an optimal value. This results from a smooth on-axis potential gradient obtained by increasing the potential gradient on the entrance/exit side of the electrostatic lens, and lens aberrations, especially, spherical aberration can be improved. In this case, the high-resistance layer 5 has two different resistances along the beam axis. However, the layer 5 may have three or more different resistance values, or the resistance may change continuously to obtain the same effect. In general, when the high-resistance layer 5 is formed so that the resistance has a positive differential coefficient, the focusing characteristics of the electrostatic lens can be improved. As another example of adjusting the on-axis electric field distribution, a method of using a larger number of high-resistance electrode substrates and applying voltages that match the lens characteristics to the wiring electrodes of these substrates is available. FIG. 4B shows an example in which four high-resistance electrode substrates 4a to 4d are used compared to the lens shown in FIG. 4A. In this example, by adjusting potentials applied to the wiring electrodes of these substrates by lens control power supplies 41a to 41c, the electric field distribution in the lens aperture 9 can be optimized. (Temperature Control) The resistor electrode lens 1 of the present invention forms an electric field in each lens aperture 9 by applying a voltage across the high-resistance layer 5, and always supplies a constant small current to the high-resistance layer 5 when it is used. Hence, by preventing temperature rise due to heating of the high-resistance layer 5 by cooling and temperature control of the lens, the resistor electrode lens 1 can stably operate. FIG. 3A shows the cooling method of the resistor electrode lens 1, and FIG. 3B is a sectional view taken along line A-Axe2x80x2 of FIG. 3A. The resistor electrode lens 1 comprises a cooling plate 33 that surrounds the lens 1, a cooling rod 34 one end of which is connected to the cooling plate 33, a temperature sensor 35 provided on the cooling plate 33, and a temperature control unit 36 to which the other end of the cooling rod 34 is connected, and can be maintained at a given temperature. Especially, the cooling plate 33 can bring the upper and lower electrode substrates with high heat conductivity into thermal contact with each other to increase temperature uniformity, thus improving cooling efficiency. The lens heating condition is determined by conditions such as the resistance of a high-resistance material used, the lens shape, the lens voltage, and the like. For example, when the high-resistance layer 5 used shown in FIG. 1 has a resistivity of 108 xcexa9cm, if the high-resistance. electrode substrate 2 has a thickness of 500 xcexcm, the lens aperture 9 has a diameter of 100 xcexcm, the high-resistance layer 5 has a film thickness of 0.2 xcexcm, the number of lenses in the multi-lens is 5,000, and the voltage applied to the wiring electrode 6 is 1,000 V, a heat quantity of around 0.6 W is generated in the resistor electrode lens. The lower limit of the resistance corresponds to a maximum heat generation quantity of 100 W of the multi-lens of this example, and a resistivity of around 106 xcexa9cm or higher. The upper limit of the resistance must be a value that can avoid charge-up of the surface of the high-resistance layer 5 due to a scattered beam or the like, and does not influence the lens characteristics of a charged beam that passes through the lens, and a high-resistance material having a resistance of around 109 xcexa9cm or less can be used. (Composition Material) As materials that form the high-resistance electrode substrate 2 in FIG. 1, various ceramics (AiN, SiC, Al2O3, BeO) or glass materials as insulators can be used. The high-resistance layer 5 can be formed as a high-resistance film with high uniformity having a thickness of 20 nm to 1,000 nm by depositing a material containing silicon carbide, nitrogen compound, or the like by sputtering. As another method of forming the high-resistance layer 5, a method using cladglass used as a channel plate material is known. With this material, the layer 5 can be formed by controlling the resistance of the inner wall of the lens aperture 9 corresponding to a channel by heating it in a hydrogen-containing atmosphere to precipitate reduced metal lead. As a material of the wiring electrode 6, a transition metal such as W, Ta, or the like, Si or silicide-based material, or the like can be used. As a material of the upper and lower electrode substrates 7 and 8, a low-resistor material is used, and a metal substrate or an impurity-doped Si substrate may be used. (Applied Apparatus) FIGS. 8 to 10 show some application examples of multi-beam lithography schemes. FIG. 8 shows an example of a multi-beam scheme using one electron source. In this example, an electron beam 88 emitted by a single electron source 81 is collimated by a condenser lens 82, and is then split into a plurality of beams by a multi-aperture and blanker 83. These beams are focused at a predetermined position on a wafer 86 using a multi-lens 84, and a stage 87 and deflector 85 are synchronously scanned, thus attaining direct write. Since the resistor electrode lenses of the present invention can be laid out at a high density per unit area, the multi-lens 84 to which these lenses are applied can generate electron beams with uniform characteristics on the basis of a beam with a relatively narrow radiation angle range emitted by an electron gun. Hence, uniformity of the irradiated beam can be improved, and the beam current density can be increased. FIG. 9 shows an example of a multi-beam lithography scheme using multi electron sources. In this example, electron beams 88 emitted by electron sources 91 pass through a multi-aperture 92 and blanker 93, and are focused by a multi-lens 84 to strike a wafer 86. In this case, direct write is done by synchronizing scan of a stage 87 and the blanker 93. Since the degree of integration of the electron beams 88 in this scheme is determined not by the size of the multi electron sources 91 but by the size of each electrostatic lens, an electron beam lithography apparatus with high degree of beam integration and high productivity can be realized by applying the resistor electrode lens of the present invention to the multi-lens 84. FIG. 10 shows an example of a multi-beam scheme using a correction optical system like that described in Japanese Patent Laid-Open No. 9-245708. In this example, an electron beam 88 emitted by a single electron source 101 is collimated by a condenser lens 102, and is split into a plurality of beams by an aperture and blanker 103. After that, the beams are reduced by a reduction electron optical system 105 via an aberration correction multi-lens 104 that corrects lens aberrations of the reduction lens 105, and strike a wafer 86. With this method, a plurality of intermediate images of the light source are formed in a direction perpendicular to the optical axis of the reduction electron optical system 105, and aberrations produced upon projecting these intermediate images onto the wafer 86 in a reduced scale via the reduction electron optical system 105, especially, field curvature and the like, can be corrected in advance. Hence, the direct write range can be broadened without decreasing the resolution of the electron optical system, and the productivity of the lithography apparatus can be improved. When the present invention is applied to such apparatus, more multi beams are formed within the effective direct write range to further improve the productivity. Wiring of the wiring electrode of the high-resistance electrode lens applied to the multi-lens is formed in correspondence with the purpose of the multi-lens in each multi-beam scheme. For example, in the examples shown in FIGS. 8 and 9, it is preferable to allow independent control of lens power in combination with height correction such as the curvature or the like of the surface of the wafer 86. On the other hand, in the example shown in FIG. 10, it is preferable to have variable lens power at the center of the beam axis in consideration of correction of field curvature. Note that the wiring electrode 6 in FIG. 1 corresponds to a wiring example with variable lens power in a linear direction. As other application examples of the present invention, the present invention is not limited to electron beam applied apparatuses such as an electron beam microscope, electron beam distance measurement apparatus, and the like, but may be applied to charged beam applied apparatuses of an ion beam and the like. In such applications, a size reduction and high processing speed of the apparatus can be achieved. As described above, according to the resistor electrode lens of the present invention, a multi-lens which can prevent charge-up due to a scattered beam in the lens aperture, can improve lens focusing characteristics of the electrostatic lens, can prevent crosstalk due to electric fields of the multi-lens, and has a high degree of integration can be realized. Also, a compact charged beam applied apparatus with high productivity can be manufactured. (Device Manufacturing Method) An embodiment of a device manufacturing method using the aforementioned electron beam lithography apparatus or charged beam applied apparatus will be explained below. FIG. 11 shows the flow in the manufacture of a microdevice (a semiconductor chip such as an IC, LSI, or the like, liquid crystal panel, CCD, thin film magnetic head, micromachine, or the like). In step 1 (circuit design), the circuit design of a semiconductor device is made. In step 2 (generate exposure control data), exposure control data of the exposure apparatus is generated based on the designed circuit pattern. In step 3 (fabricate wafer), a wafer is fabricated using materials such as silicon, and the like. Step 4 (wafer process) is called a pre-process, and an actual circuit is formed by lithography using the exposure apparatus input with the exposure control data and the wafer. The next step 5 (assembly) is called a post-process, in which semiconductor chips are assembled using the wafer obtained in step 4, and includes an assembly process (dicing, bonding), a packaging (encapsulating chips), and the like. In step 6 (inspection), inspections such as operation tests, durability tests, and the like of semiconductor devices assembled in step 5 are run. Semiconductor devices are completed via these processes, and are delivered (step 7). FIG. 12 shows the detailed flow of the wafer process. In step 11 (oxidation), the surface of the wafer is oxidized. In step 12 (CVD), an insulating film is formed on the wafer surface. In step 13 (electrode formation), electrodes are formed by deposition on the wafer. In step 14 (ion implantation), ions are implanted into the wafer. In step 15 (resist process), a photosensitive agent is applied on the wafer. In step 16 (exposure), the circuit pattern is printed on the wafer by exposure using the aforementioned exposure apparatus. In step 17 (development), the exposed wafer is developed. In step 18 (etching), a portion other than the developed resist image is removed by etching. In step 19 (remove resist), the resist film which has become unnecessary after etching is removed. By repeating these steps, multiple circuit patterns are formed on the wafer. Using the manufacturing method of this embodiment, a semiconductor device with a high degree of integration, which is hard to manufacture in the prior art, can be manufactured with low cost. As described above, according to the present invention, charge-up caused by a scattered beam in the lens aperture can be prevented by a simple arrangement, and crosstalk between electrostatic lenses can be suppressed. Hence, a multi-lens type electrostatic lens which assures a stable beam position and lens characteristics, and has a high degree of integration can be provided. By changing the resistance of a high-resistance portion, the electric field distribution in the electrostatic lens can be adjusted, and the aberration characteristics of the electrostatic lens can be improved. Furthermore, since the electrode structure is simple, the manufacturing step can be simplified, and a multi-lens with low mechanical required precision can be manufactured. Also, since a multi-lens can be prepared by laying out electrostatic lenses at high density, an electron beam lithography apparatus and charged beam applied apparatus with high productivity can be realized. Hence, the productivity in the manufacture of devices using these apparatuses can be improved. As many apparently widely different embodiments of the present invention can be made without departing from the spirit and scope thereof, it is to be understood that the invention is not limited to the specific embodiments thereof except as defined in the appended claims.
	claims	1. An apparatus comprising:a first electrode having a substantially cylindrical inner surface that has a longitudinal axis and forms at least a portion of a confining wall, wherein the confining wall at least partially encloses a confinement region;a second electrode located within a region interior to the first electrode and separated from the first electrode by at least the confinement region;at least one magnet configured to provide a magnetic field through the confinement region, at least a portion of the magnetic field in the confinement region being substantially parallel to the longitudinal axis;an inlet to the confinement region for permitting introduction of a fluid to the confinement region, the fluid containing a first reactant;a second reactant; anda control system comprising a voltage and/or current source and configured to: (a) control a potential of an electric field substantially orthogonal to the longitudinal axis, the potential being between the first electrode and the second electrode, and the potential being sufficient to produce an electrical current from the second electrode toward the first electrode;, (b) generate, from the first reactant, a weakly ionized plasma of ions and neutrals; and (c) produce a Lorentz force resulting from the electric field and the magnetic field that induces azimuthal rotation of the ions around the longitudinal axis, the azimuthal rotation of the ions imparting azimuthal rotation to neutrals of the first reactant, and promoting repeated collisions between one or both of the ions and the neutrals with the second reactant; wherein, during operation:the repeated collisions produce an interaction between the neutrals and the second reactant that produces a product having a nuclear mass that is different from a nuclear mass of any of the nuclei of the neutrals and the second reactant, and,a mole fraction of the ions to the neutrals in the weakly ionized plasma is in the range of about 0.0001% to about 1%. 2. The apparatus of claim 1, wherein the at least one magnet comprises two permanent magnets separated from one another by at least the confinement region and in the direction of the longitudinal axis. 3. The apparatus of claim 1, wherein the at least one magnet comprises two permanent magnets radially separated from one another by at least the confinement region. 4. The apparatus of claim 1, wherein at least a portion of the confining wall is separable from the remainder of apparatus to allow replacement. 5. The apparatus of claim 1, wherein the confining wall comprises a refractory metal and/or a stainless steel. 6. The apparatus of claim 1, wherein the interaction is a fusion reaction. 7. The apparatus of claim 6, wherein the fusion reaction is aneutronic. 8. The apparatus of claim 1, wherein the second reactant comprises boron-11. 9. The apparatus of claim 1, wherein the neutrals comprise one or more of neutral hydrogen, deuterium, and tritium. 10. The apparatus of claim 1, further comprising one or more electron emitters configured to emit, during operation, electrons into a region adjacent to the confining wall. 11. The apparatus of claim 10, wherein the electron emitters comprise boron or a boron-containing material. 12. The apparatus of claim 1, wherein, during operation, the confinement region proximate the confining wall comprises an electron rich region having an excess of electrons over positively charged particles of at least about 106/cm3. 13. The apparatus of claim 12, wherein, during operation, the electron rich region has an electric field strength of at least about 106 V/m. 14. The apparatus of claim 12, wherein, during operation, the neutrals in the electron rich region have an energy of, on average, of between about 0.1 eV and 2 eV. 15. A method comprising:introducing a fluid to a confinement region through an inlet to the confinement region, the fluid containing a first reactant;applying an electrical potential difference between a first electrode and a second electrode to produce an electrical current from the second electrode toward the first electrode, wherein:the first electrode has a substantially cylindrical inner surface that has a longitudinal axis and forms at least a portion of a confining wall that at least partially encloses the confinement region;the second electrode is located within a region interior to the first electrode and separated from the first electrode by at least the confinement region; andat least one magnet is configured to provide a magnetic field through the confinement region, at least a portion of the magnetic field in the confinement region being substantially parallel to the longitudinal axis; andoperating a control system comprising a voltage and/or current source so as to: (a) control a potential of an electric field substantially orthogonal to the longitudinal axis, the potential being between the first electrode and the second electrode, and the potential being sufficient to produce an electrical current from the second electrode toward the first electrode; (b) generate, from the first reactant, a weakly ionized plasma of ions and neutrals; and (c) produce a Lorentz force resulting from the electric field and the magnetic field that induces azimuthal rotation of the ions around the longitudinal axis, the azimuthal rotation of the ions imparting azimuthal rotation to neutrals of the first reactant, and promoting repeated collisions between one or both of the ions and the neutrals with a second reactant; wherein, during operation:the repeated collisions produce an interaction between the neutrals and the second reactant that produces a product having a nuclear mass that is different from a nuclear mass of any of the nuclei of the neutrals and the second reactant, and,a mole fraction of the ions to the neutrals in the weakly ionized plasma is in the range of about 0.0001% to about 1%. 16. The method of claim 15, wherein the at least one magnet comprises two permanent magnets separated from one another by at least the confinement region and in the direction of the axis of the substantially cylindrical inner surface of the first electrode. 17. The method of claim 15, wherein the at least one magnet comprises two permanent magnets radially separated from one another by at least the confinement region. 18. The method of claim 15, further comprising separating at least a portion of the confinement wall from the remainder of apparatus to allow replacement. 19. The method of claim 15, wherein the interaction is a fusion reaction. 20. The method of claim 19, wherein the fusion reaction is aneutronic. 21. The method of claim 15, wherein the second reactant comprises boron-11. 22. The method of claim 15, wherein the neutrals comprise one or more of neutral hydrogen, deuterium, and tritium. 23. The method of claim 15, wherein, during operation, the confinement region proximate the confining wall comprises an electron rich region having an excess of electrons over positively charged particles of at least about 106/cm3. 24. The method of claim 15, wherein one or more electron emitters are configured to emit, during operation, electrons into a region adjacent to the confining wall. 25. The method of claim 24, wherein the electron emitters comprise boron or a boron-containing material. 26. The method of claim 23, wherein, during operation, the electron rich region has an electric field strength of at least about 106 V/m. 27. The method of claim 26, wherein, during operation, the neutrals in the electron rich region have an energy of, on average, of between about 0.1 eV and 2 eV.
055966187	claims	1. An exposure apparatus, comprising: a holder for holding a substrate to be exposed; and a movable shutter movable across a path of exposure light, having an intensity distribution in a predetermined direction, and in a direction intersecting the predetermined direction; wherein said movable shutter has an edge with a protruded portion being protruded in the movement direction and having a shape and size determined on the bases of the intensity distribution of the exposure light. moving a movable shutter across a path of exposure light, having an intensity distribution in a predetermined direction, and in a direction intersecting the predetermined direction; wherein the movable shutter has an edge with a protruded portion being protruded in the movement direction and having a shape and size determined on the basis of the intensity distribution of the exposure light; and projecting, onto a substrate to be exposed, the exposure light passing through the movable shutter. a holder for holding a wafer to which a pattern is to be transferred by exposure; a movable shutter, movable across a path of an exposure beam having an intensity distribution, for adjusting the exposure so as to provide a uniform exposure amount within a region of the wafer to which the pattern is to be transferred; and a mechanism for changing an edge shape of said movable shutter. a holder for holding a wafer to which a pattern is to be transferred by exposure; shielding means, having an opening for partially blocking synchrotron radiation, for adjusting the exposure so as to provide a uniform exposure amount within a region of the wafer to which the pattern is to be transferred; and a mechanism for changing an edge shape of the opening of said shielding means. 2. An apparatus according to claim 1, wherein the exposure light comprises synchrotron radiation light. 3. An apparatus according to claim 2, wherein the exposure light comprises sheet-like synchrotron radiation light expanded by a mirror in a particular direction, and wherein the exposure light has a predetermined intensity distribution with respect to the particular direction. 4. An apparatus according to claim 1, wherein said movable shutter moves substantially at a constant speed, within an exposure region. 5. An apparatus according to claim 1, wherein said movable shutter is arranged to change at least one of the shape and size of the protruded portion. 6. An apparatus according to claim 1, further comprising detecting means for detecting the intensity distribution of the exposure light. 7. An apparatus according to claim 6, further comprising means for changing at least one of the shape and size of the protruded portion of said movable shutter in response to an output of said detecting means. 8. An apparatus according to claim 6, wherein said detecting means is movable as a Unit with said movable shutter. 9. A device manufacturing method, comprising the steps of: 10. A method according to claim 9, wherein the substrate comprises a wafer and wherein a pattern of a mask is transferred to the wafer through said projection of the exposure light. 11. An exposure apparatus for manufacturing a semiconductor device, said apparatus comprising: 12. An apparatus according to claim 11, wherein the exposure beam comprises synchrotron radiation. 13. An exposure apparatus for manufacturing a semiconductor device, said apparatus comprising:
	summary
	claims	1. A method of optimising the output of a sensor for indicating the relative location of a metallic object, the sensor being of the type having a primary electromagnetic coil arranged to generate a time varying magnetic field; and a secondary electromagnetic coil arranged to detect the time varying magnetic field as affected, directly or indirectly, by the object and to output, on the basis of the detected time varying magnetic field, a signal indicative of the relative location of the object, the method including the steps of:supplying the primary coil with an alternating current to result in the generated time varying magnetic field;locating the object in a first position and recording the signal output by the secondary electromagnetic coil for a range of respective frequencies of the supplied alternating current;locating the object in a second position and recording the signal output by the secondary electromagnetic coil for the range of respective frequencies of the supplied alternating current;calculating, for each of the respective frequencies, a value for the span to offset ratio of the measured signals on the basis of the respective signals measured for the object in the first and second positions, the calculating step including, for each respective frequencycalculating the difference between the amplitudes of the signals measured for the object in the first and second positions, anddividing the difference by the amplitude of the signal measured for the object in the second position; anddetermining the frequency of the supplied alternating current which provides the maximum span to offset ratio on the basis of the calculations. 2. A method according to claim 1, whereinwhen the object is in the first position, the output from the secondary coil is a maximum; and/orwhen the object is in the second position, the output from the secondary coil is a minimum. 3. A method according to claim 1, wherein at least one of the primary and secondary electromagnetic coils of the sensor is wound about a core body formed of a material having the same conductivity and/or magnetic permeability as the object. 4. A method according to claim 1 wherein the primary and secondary coils of the sensor are arranged coaxially. 5. A method according to claim 1, wherein the primary coil of the sensor is wound about a core body formed of a material having the same conductivity and/or magnetic permeability as the object. 6. A method according to claim 1, wherein the secondary coil of the sensor is wound about a core body formed of a material having the same conductivity and/or magnetic permeability as the object. 7. A method according to claim 1, wherein the primary and secondary coils of the sensor are each wound about the same core body formed of a material having the same conductivity and/or magnetic permeability as the object. 8. A method according to claim 1, wherein the primary coils of the sensor are mutually arranged in electrical series; and/or wherein the secondary coils of the sensor are separately mutually arranged in electrical series. 9. A method according to claim 1 wherein the primary and/or secondary coils of the sensor are formed of an alloy comprising 86% copper, 12% Manganese and 2% Nickel. 10. A method according to claim 1 wherein a core body of the sensor is formed of a material having the same conductivity and/or magnetic permeability as the object. 11. A method according to claim 1 wherein a core body of the sensor is formed of the same material as the object. 12. A method according to claim 1, wherein the metallic object is attached to a nuclear reactor control rod. 13. A method according to claim 1, wherein the sensor is positioned within a metallic tube and the metallic object is arranged to move relative to the tube between a position of minimum overlap and a position of maximum overlap of the tube and the object. 14. A method according to claim 13, wherein the first position is a position where there is minimum overlap between the tube and the object. 15. A method according to claim 13, wherein the second position is a position where there is maximum overlap between the tube and the object. 16. A method according to claim 1, wherein the sensor includes a plurality of primary electromagnetic coils. 17. A method assembly according to claim 16, wherein the sensor includes a plurality of secondary electromagnetic coils. 18. A method assembly according to claim 17, wherein the plurality of primary and secondary coils are arranged in a mutually alternating sequence of primary and secondary coils.
	summary
	summary
	description	The present invention will hereinafter be described specifically by the following examples. However, the present invention is not limited by these examples. A four-necked flask was equipped with a stirrer, a thermometer and a condenser and charged with 1-methoxy-2-propanol (270 g; 3.0 mol) and toluene (400 g) as a solvent, and the resultant mixture was cooled to 5xc2x0 C. with stirring. Phosphorus pentoxide (142 g; 1.0 mol) was then gradually added to the resultant solution while keeping the temperature of the solution at 5 to 10xc2x0 C. Thereafter, the temperature of the solution was gradually raised to react 1-methoxy-2-propanol with phosphorus pentoxide under conditions of 60xc2x0 C. and 6 hours. Water (20 g) was added to the resultant reaction mixture, and the resultant mixture was stirred at 80xc2x0 C. for 2 hours. After the condenser installed in the four-necked flask was replaced by a distiller, toluene and water contained in the reaction mixture were removed by the distiller, thereby obtaining a liquid reaction product (390 g). With respect to the reaction product thus obtained, spectroscopic analysis was performed by an infrared absorption spectrum. As a result, it was confirmed that the reaction product contains a phosphate compound represented by the formula (m) and a phosphate compound represented by the formula (n). The infrared absorption curve of this reaction product is illustrated in FIG. 1. With respect to the reaction product thus obtained, the compositions and yield of the phosphate compounds were calculated out in the following manner. [Composition of Phosphate Compounds] An Autotitrator COMTITE-101 manufactured by Hiranuma Sangyo K. K. was used to conduct neutralization titration of the reaction product, and the contents of the phosphate compound represented by the formula (m) and the phosphate compound represented by the formula (n) were respectively calculated out from the titers at the resultant first inflection point and second inflection point. The results are shown in Table 1. [Yield] Concentrated nitric acid and perchloric acid were added to the reaction product and the reaction product was decomposed under heat. After distilled water was added to the decomposition product to dilute it, nitric acid, a 0.25% aqueous solution of ammonium vanadate and a 5% aqueous solution of ammonium molybdate were added to the resultant solution to develop a color, thereby measuring an absorbance at a wavelength of 440 nm by means of a spectrophotometer to find a concentration (% by weight) of phosphorus in the reaction products based on the absorbance of a standard solution of phosphorus. The yield was calculated out from this concentration of phosphorus and a concentration (% by weight) of phosphorus in the phosphorus compound used. The result is shown in Table 1. A four-necked flask was equipped with a stirrer, a thermometer, a condenser to which a water scrubber had been connected, and a dropping funnel, and charged with phosphorus oxychloride (153 g; 1.0 mol), titanium tetrachloride (4.6 g) as a catalyst and toluene (180 g) as a solvent, and the resultant mixture was cooled to 5xc2x0 C. with stirring. After 1-methoxy-2-propanol (180 g; 2.0 mol) was added to the resultant solution, triethylamine (202 g; 2.0 mol) was added to the solution over 2 hours while keeping the temperature of the solution at 5 to 15xc2x0 C. The temperature of the solution was gradually raised to react 1-methoxy-2-propanol with phosphorus oxychloride under conditions of 50xc2x0 C. and 2 hours. Water (200 g) was then added to the resultant reaction mixture and hydrolysis of the reaction product was conducted under conditions of 50xc2x0 C. and 1 hour. The resultant reaction mixture was left at rest, thereby separating the reaction mixture into a toluene layer and a water layer. Thereafter, the toluene solution was recovered. Toluene (100 g) was then added to the residual aqueous solution to conduct an extraction treatment of the reaction product contained in the aqueous solution, and a toluene solution was recovered. This process was conducted repeatedly 3 times, thereby recovering the toluene solution in an amount of 600 g in total. After the condenser installed in the four-necked flask was replaced by a distiller, a treatment for removing toluene and the like from the toluene solution was conducted by this distiller, thereby obtaining a liquid reaction product (165 g). With respect to the reaction product thus obtained, spectroscopic analysis was performed by an infrared absorption spectrum. As a result, it was confirmed that the reaction product contains a phosphate compound represented by the formula (m) and a phosphate compound represented by the formula (n). The infrared absorption curve of this reaction product is illustrated in FIG. 2. With respect to the reaction product thus obtained, the composition and yield of the phosphate compounds were calculated out in the same manner as in Example 1. The results are shown in Table 1. (1) Preparation of Phosphonate Compound A four-necked flask was equipped with a stirrer, a thermometer, a condenser to which a water scrubber had been connected, and a dropping funnel, and charged with phosphorus trichloride (275 g; 2.0 mol) and hexane (200 g) as a solvent, and the resultant mixture was heated to 50xc2x0 C. 1-Methoxy-2-propanol (540 g; 6.0 mol) was then added to the resultant solution over 2 hours while keeping the temperature of the solution at 50 to 70xc2x0 C. Hydrogen chloride generated upon the addition of 1-methoxy-2-propanol in the above-described process was introduced into the water scrubber to recover it. After completion of the addition of 1-methoxy-2-propanol, the interior of the four-necked flask was sucked at 60xc2x0 C. for 1 hour under a reduced pressure of 500 mmHg, thereby conducting a treatment for removing remaining hydrogen chloride. After the condenser installed in the four-necked flask was replaced by a distiller, a treatment for removing hexane and 1-methoxy-2-chloropropane, which was a reaction by-product, in the reaction mixture was conducted by this distiller. The residue was further distilled under reduced pressure, and a distillate at 119.0 to 125.0xc2x0 C. under 3 mmHg was recovered, thereby obtaining a liquid product (398 g). This liquid product was analyzed by gas chromatography. As a result, the purity (calculated out by an area ratio in a chart) of the bis(2-methoxy-1-methylethyl) hydrogen-phosphonate was 96.3%. (2) Preparation of Phosphate Compound A four-necked flask was equipped with a stirrer, a thermometer, a condenser to which a 5% aqueous sodium hydroxide scrubber had been connected, and a dip tube for introducing chlorine gas, and charged with the above-obtained liquid product (226 g; about 1.0 mol as bis(2-methoxy-1-methylethyl) hydrogenphosphonate), and the contents were cooled to 10xc2x0 C. Chlorine gas was blown into bis(2-methoxy-1-methylethyl) hydrogenphosphonate while the temperature thereof was kept at 10 to 20xc2x0 C., and the introduction of chlorine gas was continued until the solution was slightly colored yellow. Thereafter, the interior of the four-necked flask was sucked at 25xc2x0 C. under a reduced pressure of 15 mmHg, thereby conducting a treatment for removing excess chlorine gas and hydrogen chloride, which was a reaction by-product, to obtain a liquid product (263 g). This liquid product was analyzed by gas chromatography. As a result, the purity (calculated out by an area ratio in a chart) of the bis(2-methoxy-1-methylethyl) phosphorochloridate was 92.4%. A concentration of chlorine in the liquid product was measured in accordance with xe2x80x9cDetermination Method of Chloride Ion by Silver Nitrate Standard Solutionxe2x80x9d described in xe2x80x9cExperimental Methods of Analytical Chemistryxe2x80x9d (published by Kagakudojin K. K.). As a result, the concentration of chlorine was 14.3%. Water (90 g; 5.0 mol) was added to the resultant liquid product, and the temperature of this solution was gradually raised to conduct hydrolysis of bis(2-methoxy-1-methylethyl) phosphorochloridate under conditions of 40xc2x0 C. and 2 hours. After the condenser installed in the four-necked flask was replaced by a distiller, a treatment for removing water from the reaction mixture was conducted by this distiller, thereby obtaining a reaction product (234 g). With respect to the reaction product thus obtained, spectroscopic analysis was performed by an infrared absorption spectrum. As a result, it was confirmed that the reaction product contains a phosphate compound represented by the formula (m) and a phosphate compound represented by the formula (n). The infrared absorption curve of this reaction product is illustrated in FIG. 3. With respect to the reaction product thus obtained, the composition and yield of the phosphate compounds were calculated out in the same manner as in Example 1. The results are shown in Table 1. (1) Preparation of Phosphonate Compound A four-necked flask was equipped with a stirrer, a thermometer, a condenser to which a water scrubber had been connected, and a dropping funnel, and charged with phosphorus trichloride (137.5 g; 1.0 mol) and hexane (300 g) as a solvent, and the resultant mixture was heated to 50xc2x0 C. Dipropylene glycol monomethyl ether (444 g; 3.0 mol) was then added to the resultant solution over 2 hours while keeping the temperature of the solution at 50 to 70xc2x0 C. Hydrogen chloride generated upon the addition of dipropylene glycol monomethyl ether in the above-described process was introduced into the water scrubber to recover it. After completion of the addition of dipropylene glycol monomethyl ether, the interior of the four-necked flask was sucked at 60xc2x0 C. for 3 hours under a reduced pressure of 500 mmHg, thereby conducting a treatment for removing remaining hydrogen chloride. After the condenser installed in the four-necked flask was replaced by a distiller, a treatment for removing hexane in the reaction mixture was conducted by this distiller, thereby obtaining a liquid mixture (504 g) of a phosphonate compound of dipropylene glycol monomethyl ether and a chloride of dipropylene glycol monomethyl ether, which was a reaction by-product. This liquid mixture was analyzed by gel permeation chromatography. As a result, the purity (calculated out by an area ratio in a chart) of the phosphonate compound was 72.2%. (2) Preparation of Phosphate Compound A four-necked flask was equipped with a stirrer, a thermometer, a condenser to which a 5% aqueous sodium hydroxide scrubber had been connected, and a dip tube for introducing chlorine gas, and charged with the above-obtained liquid mixture (504 g), and the contents were cooled to 10xc2x0 C. Chlorine gas was blown into this liquid mixture while the temperature thereof was kept at 10 to 20xc2x0 C., and the introduction of chlorine gas was continued until the solution was slightly colored yellow. Thereafter, the interior of the four-necked flask was sucked at 25xc2x0 C. under a reduced pressure of 15 mmHg, thereby conducting a treatment for removing excess chlorine gas and hydrogen chloride, which was a reaction by-product, to obtain a mixture (546 g) of a phosphorochloridate of dipropylene glycol monomethyl ether and a chloride of dipropylene glycol monomethyl ether. This mixture was analyzed by gel permeation chromatography. As a result, the purity (calculated out by an area ratio in a chart) of the phosphorochloridate was 69.9%. A concentration of chlorine in the mixture was measured in the same manner as in Example 3. As a result, the concentration of chlorine was 9.3%. Water (128 g; 7.0 mol) was added to the resultant mixture, and the temperature of this solution was gradually raised to conduct hydrolysis of the phosphorochloridate of dipropylene glycol monomethyl ether under conditions of 50xc2x0 C. and 2 hours. After the condenser installed in the four-necked flask was replaced by a distiller, steam distillation was conducted by this distiller under a reduced pressure or 20 mmHg while introducing steam from the dip tube, thereby conducting a treatment for removing water and the chloride of dipropylene glycol monomethyl ether from the reaction liquid to obtain a reaction product (348 g). With respect to the reaction product thus obtained, spectroscopic analysis was performed by an infrared absorption spectrum. As a result, it was confirmed that the reaction product contains a phosphate compound represented by the formula (o) and a phosphate compound represented by the formula (p). The infrared absorption curve of this reaction product is illustrated in FIG. 4. With respect to the reaction product thus obtained, the composition and yield of the phosphate compounds were calculated out in the same manner as in Example 1. The results are shown in Table 1. A process was performed in the same manner as in Example 4 except that tripropylene glycol monomethyl ether (3.0 mol) was used in place of dipropylene glycol monomethyl ether, thereby obtaining a reaction product (424 g). With respect to the reaction product thus obtained, spectroscopic analysis was performed by an infrared absorption spectrum. As a result, it was confirmed that the reaction product contains a phosphate compound represented by the formula (q) and a phosphate compound represented by the formula (r). The infrared absorption curve of this reaction product is illustrated in FIG. 5. With respect to the reaction product thus obtained, the composition and yield of the phosphate compounds were calculated out in the same manner as in Example 1. The results are shown in Table 1. less than Preparation of Resin Composition greater than The phosphate compound (hereinafter referred to as xe2x80x9cEster Axe2x80x9d) obtained in Example 3, the phosphate compound (hereinafter referred to as xe2x80x9cEster Bxe2x80x9d) obtained in Example 4 and the phosphate compound (hereinafter referred to as xe2x80x9cEster Cxe2x80x9d) obtained in Example 5 were separately used to prepare resin compositions in the following manner. The phosphate compound and methyl methacrylate were mixed in accordance with their corresponding formulations shown in Table 2. Anhydrous copper benzoate was added to the resultant mixtures, and stirring and mixing were conducted at 60xc2x0 C. for 1 hour, thereby preparing monomer compositions. t-Butyl peroxypivalate (0.2 g) was added to the monomer compositions, and the resultant mixtures were successively heated at different temperatures of 45xc2x0 C. for 16 hours, 60xc2x0 C. for 8 hours and 90xc2x0 C. for 3 hours to polymerize the methyl methacrylate, thereby preparing resin compositions (1) to (3) containing respective near infrared ray absorbers (phosphate compound and copper ion) according to the present invention. less than Evaluation of Resin Composition greater than The resin compositions (1) to (3) thus obtained were press-molded at 200xc2x0 C., thereby obtaining blue and transparent plates having a thickness of 4 mm. With respect to the plates thus obtained, the light transmittances at wavelengths of 550 nm, 800 nm and 900 nm were measured. The plates thus obtained were subjected to a 500-hour weathering test by means of a sunshine weathermeter (black panel temperature: 63xc2x0 C., precipitated), and the light transmittances of the plates after the test were measured to investigate the plates as to whether the light transmittances were changed or not. The results are shown in Table 2. As apparent from the results shown in Table 2, it was confirmed that the resin compositions (1) to (3) containing the respective near infrared ray absorbers according to the present invention have excellent visible ray-transmitting property and performance that near infrared rays are absorbed with high efficiency, and are little in the deterioration of their near infrared ray-absorbing ability by ultraviolet rays. The compound (0.14 g) represented by the above formula (a) and the compound (0.80 g) represented by the above formula (b) as specific phosphate compounds were added into methyl methacrylate (20 g) to mix them. Anhydrous copper benzoate (1.17 g) was added to the mixture solution, and the resultant mixture was stirred at 60xc2x0 C. for 1 hour, thereby reacting the phosphate compounds with anhydrous copper benzoate to prepare a monomer composition containing the specific phosphate copper compounds. t-Butyl peroxypivalate (0.2 g) was added to the monomer composition thus obtained, and the resultant mixture was successively heated at different temperatures of 45xc2x0 C. for 16 hours, 60xc2x0 C. for 8 hours and 90xc2x0 C. for 3 hours to polymerize the methyl methacrylate, thereby preparing an acrylic resin composition. The acrylic resin composition obtained in the above-described manner was evaluated. The acrylic resin composition was press-molded at 200xc2x0 C., thereby obtaining a blue and transparent plate having a thickness of 4 mm. With respect to the plate thus obtained, the light transmittances at wavelengths of 550 nm, 800 nm and 900 nm were measured. The plate thus obtained was subjected to a 500-hour weathering test by means of a sunshine weathermeter (black panel temperature: 63xc2x0 C., precipitated), and the light transmittances of the plate after the test were measured to investigate the plate as to whether the light transmittances were changed or not. The results are shown in Table 3. The spectral transmittance curve of the plate is illustrated in FIG. 6. The compounds represented by the above formula (a) to the formula (r) were provided as specific phosphate compounds (these compounds will hereinafter be referred to as xe2x80x9cEster (a)xe2x80x9d to xe2x80x9cEster (r)xe2x80x9d, respectively) to perform a process in the same manner as in Example 6 except that the specific phosphate compounds and copper salts were used in accordance with their corresponding formulations shown in following Table 3, thereby preparing acrylic resin compositions to evaluate them. The results are shown in Table 3. A four-necked flask was equipped with a stirrer, a thermometer and a condenser and charged with Ester (n) (242 g; 1.0 mol) as the specific phosphate compound, toluene (250 g) as a solvent and copper acetate monohydrate (100 g; 0.5 mol). The temperature of the mixture was gradually raised, and the mixture was stirred at 40xc2x0 C. for 1 hour and further at 80xc2x0 C. for 3 hours, thereby reacting the specific phosphate compound with copper acetate monohydrate to obtain a blue and transparent solution. The solution was subjected to a distillation treatment to remove acetic acid formed by the reaction of the specific phosphate compound with copper acetate monohydrate and toluene, thereby obtaining a phosphate copper compound (270 g) according to the present invention. The yield was 99.0%. The thus-obtained phosphate copper compound had a structure represented by the following formula (8). The phosphate copper compound was analyzed. As a result, the content of phosphorus was 11.40% by weight (theoretical value: 11.35% by weight), and the content of copper was 11.70% by weight (theoretical value: 11.64% by weight). This compound had no clear melting point, and its decomposition temperature was 247xc2x0 C. Incidentally, the infrared absorption curve of the phosphate copper compound thus obtained is illustrated in FIG. 7. The phosphate copper compound (1 g) and Ester (n) (1.03 g) were added to methyl methacrylate (20 g), and stirring and mixing were conducted at 60xc2x0 C. for 1 hour, thereby obtaining a blue and transparent monomer composition. t-Butyl peroxypivalate (0.3 g) was added to the monomer composition thus obtained, and the resultant mixture was successively heated at different temperatures of 45xc2x0 C. for 16 hours, 60xc2x0 C. for 8 hours and 90xc2x0 C. for 3 hours to polymerize the methyl methacrylate, thereby preparing an acrylic resin composition containing the phosphate copper compound according to the present invention. The resin composition was evaluated in the same manner as in Example 6. The results are shown in Table 4. A four-necked flask was equipped with a stirrer, a thermometer and a condenser and charged with Ester (p) (358 g; 1.0 mol) as the specific phosphate compound, toluene (360 g) as a solvent and copper acetate monohydrate (100 g; 0.5 mol). The temperature of the mixture was gradually raised, and the mixture was stirred at 40xc2x0 C. for 1 hour and further at 80xc2x0 C. for 3 hours, thereby reacting the specific phosphate compound with copper acetate monohydrate to obtain a blue and transparent solution. The solution was subjected to a distillation treatment to remove acetic acid formed by the reaction of the specific phosphate compound with copper acetate monohydrate and toluene, thereby obtaining a phosphate copper compound (355 g) according to the present invention. The yield was 91.3%. The thus-obtained phosphate copper compound had a structure represented by the following formula (9) and was in the form of a jelly solid. The phosphate copper compound was analyzed. As a result, the content of phosphorus was 8.03% by weight (theoretical value: 7.96% by weight), and the content of copper was 8.20% by weight (theoretical value: 8.17% by weight). Incidentally, the infrared absorption curve of the phosphate copper compound thus obtained is illustrated in FIG. 8. The phosphate copper compound (1 g) and Ester (p) (1.08 g) were added to methyl methacrylate (20 g), and stirring and mixing were conducted at 60xc2x0 C. for 1 hour, thereby obtaining a blue and transparent methyl methacrylate solution. The thus-obtained methyl methacrylate solution was used to conduct a process in the same manner as in Example 18, thereby preparing an acrylic resin composition containing the phosphate copper compound according to the present invention to evaluate it. The results are shown in Table 4. A four-necked flask was equipped with a stirrer, a thermometer and a condenser and charged with Ester (r) (475 g; 1.0 mol) as the specific phosphate compound, toluene (480 g) as a solvent and copper acetate monohydrate (100 g; 0.5 mol). The temperature of the mixture was gradually raised, and the mixture was stirred at 40xc2x0 C. for 1 hour and further at 80xc2x0 C. for 3 hours, thereby reacting the specific phosphate compound with copper acetate monohydrate to obtain a blue and transparent solution. The solution was subjected to a distillation treatment to remove acetic acid formed by the reaction of the specific phosphate compound with copper acetate monohydrate and toluene, thereby obtaining a phosphate copper compound (455 g) according to the present invention. The yield was 90.0%. The thus-obtained phosphate copper compound had a structure represented by the following formula (10) and was in the form of a viscous liquid. The phosphate copper compound was analyzed. As a result, the content of phosphorus was 6.20% by weight (theoretical value: 6.13% by weight), and the content of copper was 6.33% by weight (theoretical value: 6.29% by weight). Incidentally, the infrared absorption curve of the phosphate copper compound thus obtained is illustrated in FIG. 9. The phosphate copper compound (1 g) and Ester (r) (1.18 g) were added to methyl methacrylate (20 g), and stirring and mixing were conducted at 60xc2x0 C. for 1 hour, thereby obtaining a blue and transparent methyl methacrylate solution. The thus-obtained methyl methacrylate solution was used to conduct a process in the same manner as in Example 18, thereby preparing an acrylic resin composition containing the phosphate copper compound according to the present invention to evaluate it. The results are shown in Table 4. Ester (c) (0.4 g) and Ester (d) (1.6 g) as the specific phosphate compounds and anhydrous copper benzoate (1.3 g) were added to toluene (20 g), and the resultant mixture was stirred and mixed at 60xc2x0 C. for 1 hour, thereby reacting the specific phosphate compounds with anhydrous copper benzoate to obtain a blue and transparent toluene solution containing the phosphate copper compounds according to the present invention. The whole amount of the toluene solution was added to polymethyl methacrylate beads (xe2x80x9cMHGAxe2x80x9d, product of Sumitomo Chemical Co., Ltd.; 40 g) to stir and mix them. Thereafter, the resultant mixture was dried under reduced pressure at 60xc2x0 C. for 24 hours to conduct a treatment for removing toluene, thereby obtaining a massive product. The massive product was ground and then kneaded for 5 minutes by rolls heated to 180xc2x0 C., thereby obtaining a blue and transparent resin composition. The resin composition was press-molded, thereby producing a plate having a thickness of 2 mm to evaluate it in the same manner as in Example 6. The results are shown in Table 4. As apparent from the results shown in Tables 3 and 4, it was confirmed that the resin compositions containing the respective phosphate copper compounds according to the present invention have excellent visible ray-transmitting property and performance that near infrared rays are absorbed with high efficiency, and are little in the deterioration of their near infrared ray-absorbing ability by ultraviolet rays. A four-necked flask was equipped with a stirrer, a thermometer and a condenser and charged with Ester (n) (242 g; 1.0 mol) as the specific phosphate compound and toluene (250 g) as a solvent. The mixture was cooled to 5xc2x0 C., and 25% aqueous sodium hydroxide (160 g, 1.0 mol as sodium hydroxide) was gradually added to the solution while keeping the temperature of the solution at 5 to 20xc2x0 C., thereby neutralizing Ester (n). Thereafter, an aqueous solution with copper (II) sulfate pentahydrate (250 g; 1 mol) dissolved in water (750 g) was added to this solution over 1 hour while keeping the temperature of the solution at 20xc2x0 C. The temperature of the solution was gradually raised to react Ester (n) with copper (II) sulfate under conditions of 80xc2x0 C. and 5 hours. After sodium sulfate and sodium copper (II) sulfate formed in the reaction mixture were removed by filtration, the reaction mixture was left at rest, thereby separating the reaction mixture into a toluene layer and a water layer to recover the toluene solution. Toluene (200 g) was then added to the residual aqueous solution to conduct an extraction treatment of the reaction product contained in the aqueous solution, and a toluene solution was recovered. This process was conducted repeatedly 3 times, thereby recovering the toluene solution in an amount of 1020 g in total. After the condenser installed in the four-necked flask was replaced by a distiller, a treatment for removing toluene and the like from the toluene solution was conducted by this distiller, thereby obtaining a reaction product (180 g). The yield was 66.0%. The thus-obtained phosphate copper compound had the structure represented by the above formula (8). The phosphate copper compound was analyzed. As a result, the content of phosphorus was 11.26% by weight (theoretical value: 11.35% by weight), and the content of copper was 11.04% by weight (theoretical value: 11.64% by weight). This compound had no clear melting point, and its decomposition temperature was 240xc2x0 C. Incidentally, the infrared absorption curve of the phosphate copper compound thus obtained is illustrated in FIG. 10. According to the phosphate compounds of the present invention, a copper ion can be dispersed in a high proportion in a synthetic resin, since they have a hydroxyl group capable of being tonically or coordinately bonded to the copper ion, and are good in compatibility with synthetic resins, for example, acrylic resins. Accordingly, the phosphate compounds according to the present invention are suitable for use as additives for resins for dispersing a copper ion in synthetic resins. According to the process of the present invention for preparing a phosphate compound, the phosphate compound can be prepared with advantages. The phosphate copper compounds according to the present invention have performance that near infrared rays are absorbed with high efficiency, and are little in the deterioration of its near infrared ray-absorbing ability by ultraviolet rays and satisfactory in compatibility with synthetic resins, for example, acrylic resins. According to the process of the present invention for preparing a phosphate copper compound, the phosphate copper compound can be prepared with advantages. According to the near infrared ray absorbers of the present invention, a copper ion, which is a near infrared ray-absorbing component, can be dispersed in a high proportion in a synthetic resin because the above-mentioned phosphate compound is contained therein, and the resin compositions, to which the near infrared ray absorber according to the present invention is added, have excellent visible ray-transmitting property and performance that near infrared rays are absorbed with high efficiency and are little in the deterioration of its near infrared ray-absorbing ability by ultraviolet rays. According to the near infrared ray absorbers of the present invention, resin compositions, which have performance that near infrared rays are absorbed with high efficiency and are little in the deterioration of its near infrared ray-absorbing ability by ultraviolet rays and excellent in visible ray-transmitting property, can be provided by containing such an absorber in a synthetic resin, since the near infrared ray absorbers comprise the phosphate copper compound as an effective ingredient. The near infrared ray-absorbing acrylic resin compositions according to the present invention have excellent visible ray-transmitting property and performance that near infrared rays are absorbed with high efficiency and are little in the deterioration of its near infrared ray-absorbing ability by ultraviolet rays.
	summary
061334987	claims	1. A method for producing phosphate ceramic products, comprising the steps of: contacting a starter oxide with a phosphate-containing solution to create phosphates of the oxide in a slurry; adding at least one reducing agent to the slurry; mixing the slurry; and allowing the slurry to set into a phosphate ceramic product. combining the waste with a starter oxide to obtain a powder; contacting the waste and powder mixture with a phosphate-containing solution to create phosphates of the oxide in a slurry; adding at least one reducing agent, such that the oxidation state of the metal anion is reduced; mixing the slurry; and allowing the slurry to set into a phosphate ceramic product, thereby chemically and physically encapsulating the waste. 2. The method according to claim 1, wherein the step of adding the reducing agent includes adding the reducing agent to the starter oxide prior to contacting the starter oxide with the phosphate containing solution. 3. The method according to claim 1, wherein the reducing agent is a stannous salt. 4. The method according to claim 1, wherein the reducing agent is tin chloride. 5. The method according to claim 1, wherein the reducing agent is selected from the group consisting of sodium monosulfide, potassium monosulfide, calcium sulfide, iron sulfide, iron sulfate, sodium thiosulfate, sulfur dioxide, sodium borohydride, hydrazine, sodium bisulfite, calcium hydroxide, sodium hydroxide, sodium carbonate, sulfuric acid, and formic acid. 6. The method according to claim 1, further comprising the step of first heating the starter oxide in a reducing atmosphere. 7. The method according to claim 1, further comprising the step of first calcining the starter oxide in a vacuum. 8. The method according to claim 1, wherein the starter oxide is haematite. 9. The method according to claim 1, wherein the starter oxide is an oxide or hydroxide of an element or compound selected from the group consisting of aluminum, calcium, iron, magnesium, titanium, and zirconium, and combinations thereof. 10. The method according to claim 1, wherein the phosphate-containing solution is comprised of a solution selected from the group consisting of phosphoric acid and acid phosphate, and further comprises an acid phosphate of a compound selected from the group consisting of aluminum, ammonium, beryllium, calcium, iron, lanthanum, magnesium, magnesium-sodium, magnesium-potassium, sodium, yttrium, zinc, and zirconium, and combinations thereof. 11. The method according to claim 1, wherein the phosphate-containing solution is phosphoric acid having a concentration in the range of between about 30 and about 50 weight percent and wherein the starter oxide is haematite. 12. The method according to claim 1, wherein the phosphate-containing solution is a diluted phosphoric acid having a concentration in the range of between about 30 and about 40 weight percent, and wherein the starter oxide is selected from the group consisting of magnetite or iron oxide. 13. The method according to claim 1, further comprising the step of partially neutralizing the phosphate-containing solution prior to the step of contacting the starter oxide with the phosphate-containing solution, thereby decreasing the rate of reaction. 14. The method according to claim 1, further comprising the step of adding boric acid to the slurry to control the rate of reaction. 15. The method according to claim 1, further comprising the step of adding a waste material to the slurry, such that the waste material is encapsulated by the phosphate ceramic product. 16. The method according to claim 1, wherein the phosphate ceramic product is a glass-crystalline ceramic. 17. The phosphate ceramic product produced by the method of claim 1. 18. A method for stabilizing waste containing at least one metal anion, in a phosphate ceramic product, comprising the steps of: 19. The method according to claim 18, wherein the step of adding the reducing agent includes adding the reducing agent to the waste prior to combining the waste with the starter oxide powder. 20. The method according to claim 18, wherein the step of adding the reducing agent includes adding the reducing agent to the waste and powder mixture prior to contacting the waste and powder mixture with the phosphate-containing solution. 21. The method according to claim 18, wherein the step of adding the reducing agent includes adding the reducing agent to the slurry. 22. The method according to claim 18, wherein the reducing agent is a stannous salt. 23. The method according to claim 18, wherein the reducing agent is tin chloride. 24. The method according to claim 18, wherein the reducing agent is selected from the group consisting of sodium monosulfide, potassium monosulfide, calcium sulfide, iron sulfide, iron sulfate, sodium thiosulfate, sulfur dioxide, sodium borohydride, hydrazine, sodium bisulfite, calcium hydroxide, sodium hydroxide, sodium carbonate, sulfuric acid, and formic acid. 25. The method according to claim 18, wherein the starter oxide is an oxide or hydroxide of an element or compound selected from the group consisting of aluminum, calcium, iron, magnesium, titanium, and zirconium, and combinations thereof. 26. The method according to claim 18, wherein the phosphate-containing solution is comprised of a solution selected from the group consisting of phosphoric acid and acid phosphate, and further comprises an acid phosphate of a compound selected from the group consisting of aluminum, ammonium, beryllium, calcium, iron, lanthanum, magnesium, magnesium-sodium, magnesium-potassium, sodium, yttrium, zinc, and zirconium, and combinations thereof. 27. The method according to claim 18, wherein the metal anion is an anion of a metal selected from the group consisting of antimony, arsenic, chromium, mercury, selenium, and technetium. 28. The method according to claim 18, wherein the metal is chromium and the reducing agent is selected from the group consisting of tin chloride and sodium monosulfide. 29. The method according to claim 18, wherein the metal is selected from the group consisting of arsenic, selenium, and technetium and the reducing agent is tin chloride. 30. The method according to claim 18, wherein the metal is mercury and the reducing agent is sodium monosulfide. 31. The method according to claim 18, wherein the metal anion is chromate and the step of adding a reducing agent to the slurry reduces the valency of chromate from +6 to +4. 32. The method according to claim 18, wherein the metal anion is technetium-99 and the step of adding a reducing agent to the slurry reduces the valency of technetium-99 from +7 to +4. 33. The method according to claim 18, wherein the metal anion is technetium-99 and the pH of the slurry is maintained at no greater than 7.
050531891	description	DESCRIPTION OF THE PREFERRED EMBODIMENT The Nuclear Reactor More particularly, there is shown in FIG. 1 a pressure vessel body 32 for a pressurized water nuclear reactor 30. A dome 34 is provided for the vessel 32 and an outer core barrel 36 is secured to and extends downwardly from a flange 38 where the dome and vessel body are secured together. The bottom of the core barrel 36 is secured to the vessel body 32 as indicated by the reference character 37. A fuel core 40 is supported within the core barrel 36 by a horizontal upper core support plate 42 and a horizontal lower core support plate 44. The core 40 comprises an array of vertically elongated fuel assemblies 41, in this case a total of 157 to 193 fuel assemblies for 3 and 4 loop plants. Each fuel assembly comprises an array (not detailed) of vertically elongated fuel rods, in this case a 17.times.17 rod array with predetermined rod locations having control rod thimbles (not shown) instead of fuel rods. Horizontal grid members (not shown) provide support against fuel rod bending at various points along the height of each fuel assembly. As subsequently considered more fully, only about 30% of the assemblies are designated for control rod operation. Thus, a total of forty-eight to sixty-one fuel assemblies are used in this case for control rod insertion. The remaining fuel assemblies are inactive from a control rod standpoint, and the control rod thimbles for those assemblies may be end plugged. A horizontal upper support plate 46 is located within the core barrel 36 at an elevation just below the flange 38. The space 47 formed between it and the upper core support plate 42 is called the outlet plenum since the reactor coolant heated in an upward pass through the fuel core 40 is collected in this space for outflow through two or more outlet nozzles 48 to the steam generator(s) (not shown). Return flow from the steam generators passes through two or more inlet nozzles 50 to a downcomer space 52 between the pressure vessel body 32 and the core barrel 36. The return flow mainly moves downwardly through the downcomer space to the bottom of the fuel core 40 where it begins another upward pass through the fuel core 40 in the spaces between the fuel rods to absorb heat generated by the nuclear reaction. As previously noted, some of the return flow is bypassed from the core 40 for flow from the top of the downcomer 52 through spray nozzles into the space 54, called the upper plenum, above the upper support plate 46. The bypass flow to the upper plenum passes through nozzles located in the flange of the core barrel 36 as indicated by arrows 56. Reactor control is provided by a plurality of control rod assemblies 60, in this instance forty-eight to sixty-one control rod assemblies as previously indicated. The control rod assemblies 60 are provided with control rod guidance support by respective guide tube assemblies 62 extending vertically through the upper head plenum 54 and the outlet plenum 47 as shown. As observed in FIG. 11, the upper support plate 46 is provided with a number of openings 64 corresponding in number and aligned with the downwardly located fuel assemblies. Respective cover plates 65 close the openings 64 located above fuel assemblies for which no control rods are provided. The control rod guide tube assemblies 62 extend through those openings 64 aligned with fuel assemblies for which control rods are provided. Generally, the guide tube assemblies 62 are bolted or otherwise secured to the upper support plate 46 such that the upper support plate 46 conventionally forms a barrier against flow from the upper head plenum 54 to the outlet plenum 47. Coolant flow does pass from the upper plenum 54 through the guide tube assemblies to the outlet plenum 47. (The purpose of the flow is to maintain a cold reservoir of water in the upper head to provide better loss of coolant safety margins for the fuel in case of an accident). Respective control rod drives mechanisms 66 are located above the vessel dome 34 in alignment with the respective control rods. Each mechanism operates a downwardly extending drive rod 68 that extends through a dome penetration 69 and is coupled to each control rod assembly to implement control over control rod position by the drive mechanisms 66. Some coolant does pass from the upper plenum 54 through the control rod drive mechanisms 66 and back to the plenum 54 to prevent overheating of the control rod mechanisms 66. Normally, only four or five preselected control rod assemblies need to be inserted into the fuel core 40 to regulate the level of power generated by the nuclear reaction. The remaining fifty or so control rod assemblies are normally fully withdrawn from the fuel core 40 during reactor operation, that is the control rod assembly itself is positioned so that its top is near the top of the guide tube assembly 62 and its bottom is just inserted into the top of the thimble tubes of the fuel assemblies 41. To shut the reactor down, all control rods are fully inserted into the fuel assemblies 41. Control Rod Guide Tube Assemblies--More Detail More detail is shown for the control rod assemblies 60 and guide tube assemblies 62 in FlGS. 2 and 3. In FIG. 2, the guide tube assembly 62 is shown aligned with a fuel assembly 41. The guide tube assembly 62 includes an upper guide tube 62U bolted to the top of the upper support plate 46 and an aligned lower guide tube subassembly 62L bolted to the bottom of the upper support plate 46. In turn the lower guide tube subassembly includes an intermediate guide tube 63A and a lower guide tube continuous section 63B that terminates at the upper core support plate 42. Guide tube support cards 70 (70A of FIG. 3A or 70B of FIG. 3B) are secured within the guide tube assemblies 62 at spaced locations along the guide tube height to provide guided support for the individual control rodlets of each control rod assembly 60. The spacing between the guide tube support cards may for example be one foot. Operating experience with nuclear power plants indicates that control rodlet wear is greatest at rod locations near the guide tube support cards 70 principally as a result of flow induced rod vibrations. Further, significantly greater rodlet wear occurs for the withdrawn control rods in the upper guide tubes 62U. Coolant flow through the guide tubes is a major contributor to rodlet vibrations that are the main cause of rodlet wear near the guide tube support cards 70. As shown in FIGS. 3 and 5, each control rod assembly 60 includes a top spider plate 80 that is coupled to the drive rod at 68. The spider 80 includes a plurality of radially extending vanes 82, each of which supports one or more control rodlets 84 that extend downwardly therefrom to the bottom of the control rod assembly. Each rodlet 84 may for example comprise a hafnium core which operates as a neutron absorber and an outer stainless steel cladding. The individual rodlets 84 extend through aligned openings or slots 85 or 86 (FIGS. 3a, 3b) in the guide tube support cards 70 or lower guide tube continuous section 63B. The guiding support relationship thus provided by the guide tube support cards 70 for the rodlets 84 accordingly is also a major source of rodlet wear under flow induced rodlet vibration. As shown in the top plan view of FIG. 4, each control rod in this case includes a total of twenty-four rodlets 84. In FIG. 6, a coding scheme is used to show the relative wear that typically occurs for the various rodlets 84. Thus, high wear H occurs for the four inside face rodlets; high/medium wear H/M occurs for the eight outer single vane rodlets; medium wear M occurs for the four inside corner rodlets; and low wear L occurs for the remaining eight outer rodlets. System Providing Guidance Support With Restricted Coolant Flow For Control Rods Generally, reduced control rod wear is achieved in accordance with the invention by the application of structure at the upper end of the upper control rod guide tube that reduces the flow of coolant from the upper plenum through the guide tube particularly when the control rod is located in its withdrawn position. With the guide tube flow thus limited, bypass flow paths are provided through the upper support plate to maintain desired flow from the upper plenum to the outlet plenum. The typical prior art structure for the upper guide tube and the control rod assembly in the vicinity of the upper end of the upper guide tube is shown in FIG. 7 with the control rod in its fully withdrawn position. Thus, the drive rod 68 extends downwardly from above and it is located within a thermal sleeve 90 having a bottom funnel 92 that terminates with some spacing above the top of the upper control rod guide tube 62U. The control rod 60 is located within the upper control rod guide tube 62U and extends further downwardly into the top of the thimble tubes of the fuel assemblies 41 (FIGURE 1). A coupling 94 is provided to support the control rod 60 from the drive rod in end-to-end relationship. Thus, the coupling 94 mechanically links the end portion 96 of the drive rod 68 to the control rod spider 80. The rod coupler 94 includes a threaded portion 98 that securely attaches to the drive rod bottom end end portion 96 within the upper guide tube 62U. It also includes a bottom end gripper 100 that securely engages the control rod spider 80. A protective sleeve 102 covers the coupler flexures 95 from the bottom portion 100 to a point just short of the gripper portion 98 as indicated by the reference character 104. In the fully withdrawn control rod position, the drive rod thus extends downwardly into the upper control rod guide tube 62U where it is coupled to the control rod spider and the control rod extends further downwardly therefrom. When the control rod drive mechanism is operated to lower the control rod into the fuel core, the drive rod/control rod subassembly is lowered through the guide tube structure until the desired control rod position is reached. The upper guide tube 62U includes a cylindrical side wall 106 and a top enclosure plate 108 that extends over the open upper end of the tube wall 104. A hole 110 is located in the top plate 108 for the drive and control rod assembly. The hole 110 is large enough to provide a clearance space 112 between the rod assembly and the body of the top guide tube plate 108 both to provide for inlet coolant flow to the upper guide tube 62U from the upper plenum and to permit unobstructed upward and downward movement of the drive and control rod assembly. There is shown in FIG. 8 improved structure that restricts coolant flow into the upper inlet end of the upper control guide tube in accordance with a preferred embodiment of the invention. The FIG. 8 embodiment is preferred since it achieves desired results yet can be applied economically as a retrofit to provide an improvement update for preexisting field installations. Where the prior art and the improvement structure involve like structural elements, corresponding reference characters are employed. To restrict inlet coolant flow, a flow restrictor 120 is bolted or otherwise secured to top plate 122 as indicated by the reference character 124. Apart from the provision for flow restrictor securance, the top plate 122 is otherwise like the prior top plate 108. Bolt 126 is employed to prevent excessive twisting of the spider during scram. The flow restrictor 120 is formed with a central opening 126 preferably having a diameter equal to the hole 110 in the top guide tube plate 122. A flow contraction at 128 and a flow expansion into enlarged space 130 within the flow restrictor 120 provides flow resistance in addition to that provided by a top plate flow contraction at 132 and expansion into enlarged space 134 within the upper guide tube 62U. Additional flow resistance is provided by a reduction in the flow cross-section of the clearance space 136 between the drive rod 68 and the flow restrictor 120 and the clearance space 138 between the drive rod 68 and the top plate 122. Preferably, the reduced flow area is achieved by providing drive/control rod coupling 140 with a protective sleeve 142C similar to the prior art protective sleeve 102 with an extended flow restrictor sleeve 142 that is positioned over the existing coupling 140 and drive rod 68. This modified protective sleeve with flow restrictor sleeve extension 142 is attached by threading to the coupler 140 in the same manner as the prior art protective sleeve. The flow restrictor sleeve 142 extends upwardly over the drive rod 68 through the opening 110 in the guide tube top plate 122 and the opening 126 in the flow restrictor 120. The upper end of the flow restrictor sleeve 142 can be welded or pinned to the drive rod 68. Additional flow restriction can be achieved by providing a flow restrictor with additional contraction/expansion structure within the control rod guide tube 62U. However, it is preferable that added flow restriction be achieved by flow area reduction as already described since additional flow restrictor structure within the guide tube requires that the guide tube top plate be cut away to permit insertion of the flow restrictor into the guide tube. In retrofit applications, such cutting and any drilling has to be performed under water. In any case, achievable increases in the loss coefficient would not likely be enough to produce desired flow reduction through the control rod guide tube 62U, and flow area reduction is also needed as already described. Overall, coolant flow is limited by increased hydraulic resistance K/Asquared. That is, the loss coefficient K is increased by additional contraction/expansion and the flow area A is reduced by the sleeve 142. The drive rod clearance spaces 136 and 138 with the control rod in its illustrated fully withdrawn position are thus the respective gaps between the sleeve 142 and the flow restrictor 120 and the top guide tube plate 122. With this arrangement, the original drive rod clearance space 112 shown in FIG. 7 is available for drive rod movement when the control rod is lowered from its fully withdrawn position and the top end 145 of the sleeve 142 has dropped below the top guide tube plate 122. When the control rod is lowered and the original rod clearance is applicable, the flow resistance is significantly reduced since the guide tube inlet flow area is greater. However, as previously noted the greatest control rod wear occurs with the control rod fully withdrawn for reasons including the fact that the great majority of the control rods are fully withdrawn for most of the reactor operating time. Upper Support Plate Flow Redistribution In implementing the invention, it is preferable that provision additionally be made for redistributing the coolant flow restricted from entry to the control rod guide tubes 62U, i.e. a bypass flow from the upper plenum 54 to the outlet plenum 47 is preferably provided to make up the flow reduction through the control rod guide tubes 62U. The unused control rod guide tube openings 64 in the upper support plate 46 are preferably used for this purpose, especially in retrofit applications of the invention. Thus, the cover plates 65 are removed from selected control rod guide tube openings 64 and in their places respective orifice plates 150 (FIGURES 11a and 11b) are secured. The total flow area through openings 152 in the orifice plates 150 provides the desired total bypass flow from the upper plenum 54 to the outlet plenum 47. The radial pressure profile across the outlet plenum is employed in each plant in determining the location of the orifice plates and the size of the orifice openings so as to provide a total inter-plenum flow equivalent to the total flow prior to a guide tube flow restriction retrofit. In other applications of the invention, special openings may be provided in the upper support plate 46, or flow openings 67 may be provided in lower flange structure or the like provided at the bottom of the upper control rod guide tube 62U (see FIGURE 11c). Other Embodiments Of The Invention Selected additional embodiments of the invention in FIGS. 9 and 12-16. As shown in FIG. 9, flow restriction can be produced instead by a drive rod 68A having an end portion 68B with an enlarged diameter thereby reducing the coolant flow area within the flow restrictor and upper guide tube plate opening. In this case, a protective sleeve 102A like the prior art sleeve 102 is all that is required for the drive rod/control rod coupling. The FIG. 9 embodiment would be most applicable to new nuclear power plants. In a variation (not shown) of the FIG. 8 embodiment, a flow restrictor sleeve dropped over the drive rod 68 until its lower end portion covers the upper part of the rod coupling and abuts the protective sleeve. The sleeve lower portion is then crimped against wrench flats on the coupling (see FIG. 12 at 68D). A modified flow restrictor 160 (FIG. 13) can be employed for 4 loop nuclear plants with guide tubes which characteristically have more space between top plate 122A of the upper control rod guide tube 62U and the thermal sleeve guide 92. In this case, the flow restrictor 160 has structure 160A that extends into the enlarged space and helps prevent drive rod vibrations otherwise caused by upper plenum cross-flow. The structure 160A further provides three cascaded contraction/expansion losses 161A, 161B, and 161C. If desired, the flow restrictor 160 may be further structured to include a lower extension 162 having structure that provides three additional cascaded contraction/expansion losses 163A, 163B, and 163C. However, the lower extension requires that the top upper guide tube plate be cut away as previously described. A flow restrictor 166 in FIG. 12 is similar to the flow restrictor 160 of FIG. 13, except that it is designed for standard plants like the one applicable to FIG. 8 where standard spacing is provided between the thermal sleeve and the top guide tube plate. Thus, the flow restrictor 166 has a single contraction/expansion loss above top plate 122E and is structured like the flow restrictor 160 with three cascaded contraction/expansion losses below the top plate 122E. In FIG. 14, see attached FIG. 14, another modified flow restrictor 170 is shown for use in plants with rapid refueling feature which have a relatively longer core (14 feet versus 12 feet) and longer control rod assembly. To operate rapid refueling, the control rod assembly must be able to step up 7.5 inches above the normal 258 step position. This restricts the length of the B restrictor sleeve 176 to prevent interference with the thermal sleeve funnel 92. Because of reduced space 172, the flow restrictor 170 has a portion 174 that extends through an enlarged opening in the top guide tube plate 122. A short protective sleeve 176 is attached by threading to the original pickup button 177. The restrictor sleeve 176 length is designed to engage any one of the flow restrictor 170 contractions over a range of drive rod 176 positions to provide flow area reduction over such range if repositioning is employed to spread the control rod wear over several positions to reduce wear depth. The flow restrictor can be modified to be a flow limiter, i.e. a device that blocks coolant flow through the control rod guide tube 62, and further to accommodate changes in the distance between the thermal sleeve and the upper guide tube 62 caused by thermal growths and pressure bowing of the vessel dome or the upper support plate 46. In FIG. 15, a bellow spring 180 is seated against top plate 122C to preload flow blocker 182 against the thermal sleeve 92. A spherical seat 184 maintains ring contact around the thermal sleeve 92 regardless of any cone tilt of the thermal sleeve 92. In FIG. 16, a coil spring 190 is seated against top plate 122D to preload the flow blocker 182 against the thermal sleeve 92. Generally, the coil spring 190 is more flexible than the bellow spring 180 since it provides blocker/sleeve contact control over a wider range of axial distances.
	summary
039403130	claims	1. A nuclear reactor fuel rod comprising a nonmagnetic tubular jacket having an end portion, an end cap closing said end portion, nuclear fuel in said jacket, said fuel and end cap defining therebetween a space inside of said jacket, a capsule-like housing positioned in said space, said housing being made of non-magnetic material and having at least one hole formed therethrough, a mixture of aluminum powder and a water-soluble mercury compound inside of said capsule-like housing, and ferromagnetic material inside of said capsule-like housing, said housing internally forming a space into which said mixture and said material can move. 2. The fuel rod of claim 1 in which said mixture is in the form of a compact body. 3. The fuel rod of claim 2 in which said ferromagnetic material is in the form of a powdered component of said body. 4. The fuel rod of claim 2 in which said ferromagnetic material is in the form of a pellet abutting said body. 5. The fuel rod of claim 2 in which said capsulelike housing contains a cup-like container, said body forms a covering lid for said container and said ferro-magnetic material is contained in said container.
041773853	description	DESCRIPTION OF THE PREFERRED EMBODIMENT FIGS. 1 and 2 are general arrangement view, wherein an egg crate structure is formed of stainless steel plates 10. These plates extending through the entire height of the rack must be generally equal to or greater in length than the active length of the fuel assemblies to be stored and must be coterminous therewith when the assemblies are stored within the rack. Support bars 12 pass through the lower portion of the rack for support of the fuel assemblies to be stored. This egg crate structure forms vertically extending rectangular openings 14 through 22. Fuel of low effective enrichment may be stored in each of these openings. Fuel of a higher enrichment (typically about 3.5 to 4.0 weight percent U 235) which is normally expected to be stored, may be safely stored in a checkerboard pattern using openings 15, 17, 19, 21 etc. With practical center to center spacing, this will normally provide safe storage of the highest enriched fuel expected to be stored. Since only one half the openings are used, the capacity of the rack with such a storage mode is limited to one half the possible capacity. As the capacity of the storage rack in this mode is approached, it may be desirable to increase the capacity. This is accomplished by adding inserts 24 of stainless steel or some other neutron absorbing material as illustrated in FIG. 3. These inserts are formed of two plates 26 and 28 which are parallel to adjacent sides of the opening and are spaced therefrom a predetermined distance. The appropriate distances and fuel enrichment capability are calculated by considering the total mass of the fuel to be stored, the plates 10 of the egg crate structure, the inserts 26, and the water gap 30 between the sides of the boxes and the plates. Each insert has a bent portion, or spacer 32 at each end and a spacer piece 33 at the intervening corner. These cooperate to maintain the insert properly spaced within the opening and to maintain the predetermined space of the water gap 30. As best seen in FIGS. 2 and 4 rectangular openings 34 are supplied in plates 10 of the egg crate structure. The inserts have an upwardly extending tab 36 which is bent so as to lock into the opening 34. Tab extensions 38 are welded to the tab and prevent the tab from passing through the opening 34. The insert is supported on plate 12 and the tabs function to maintain the insert in position and retain the insert in position when the stored fuel assemblies are withdrawn. When fuel of still higher enrichment must be stored there are several options available, each of which involves the use of a high neutron absorbing plate such as material containing boron 10 or hafnium. Referring to FIG. 5, poison plates 40 and 42 are installed in the water gap between the insert 24 and the side plates 10 of the opening. In the particular construction illustrated these plates may be inseted by springing back tab 38 and inserting the plate. They may be supported either from the tab itself or on an additional support member added to the egg crate structure at the bottom. The enrichment of fuel which may be stored at this time must be calculated taking into consideration the poison plate as well as the times discussed with respect to the flux trap principle arrangement. An alternate method of storing additionally enriched fuel is illustrated in FIG. 6 wherein the poison plates 40 and 42 of FIG. 5 are removed along with the insert 24. A rectangular storage box 50 of stainless steel or a higher neutron absorber poison material, which is commercially available, may be installed in the openings and supported on bars 12. The enrichment of fuel which may be stored in this mode must again be calculated by nuclear physics principles according to the physical structure which exists.
	abstract	Method for controlling the amount of metal atoms deposited into an oxide layer present on a metal surface, which metal atoms increase the corrosion resistance of metal when present in the oxide film, wherein the metal surface is submerged in water at a selected temperature within the range of about to 200xc2x0 to 550xc2x0 F.; and a solution of a compound containing the metal which increases the corrosion resistance of the metal surface when present in the oxide film is injected into the water. The compound decomposes at the selected temperature to release atoms of the metal which incorporate in the oxide film at a desired loading.
	abstract	Arranging an application apparatus, an exposure apparatus, a developing apparatus, and similar apparatus together within a large yellow room makes it difficult to change a layout in association with a change of a recipe and to streamline a layout area and similar parameter. A yellow room system includes: a plurality of portable unit process apparatuses 50 that each have the same standardized outer shape and include a yellow room configured to shield a exposure light to a photosensitive material formed on a workpiece; conveyance containers 11 and 25 that convey the workpiece between the unit process apparatuses and itself is formed as the yellow room; and a light-shielding coupling structure that couples the unit process apparatus, which is formed on a docking port 56 disposed in the upper portion of a front chamber 80 of the unit process apparatus 50, and the conveyance containers together.
050948059	description	Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, there is seen a 17.times.17 fuel assembly 1, which includes a fuel assembly head 2 with holding-down springs 3, a fuel assembly foot 4, a number of guide tubes 5 (in this case the number n equals 24), an instrumentation tube 5a, square grid-like spacers 6, and round fuel rods 7, 8, 9 and 10, which are filled with fuel in the form of cylindrical pellets. Edge fuel rods located at the edge or edge leg 11 shown in FIG. 2, are identified by reference numeral 10. The fuel rods 7, 8, 9 and 10 are conventionally disposed in a grid configuration with a strictly square cross section, or in other words in a specified equal distribution. As will become clear below, in at least one embodiment of the invention, a departure from this strictly square grid configuration is made. The possible structural difference between the fuel rods 7-10 will be explained later. The fuel rods 7 may be conventional standard fuel rods, for example. As a rule, in a reactor core, a given number of fuel assemblies 1 which are identical to one another, are provided. The fuel assemblies are disposed in a reactor pressure vessel and a coolant K flows through them from bottom to top, as indicated by an arrow. The fuel rods 7, 8, 9, 10 of the fuel assembly 1 are retained in a support structure, which is formed of the fuel assembly head 2, provided with the four holding-down springs 3, the fuel assembly foot 4, the guide tubes 5 inbetween four non-illustrated control rods and the central instrumentation tube 5a, depending on the fuel assembly type. The grid-like spacers 6 which have a square spacer cross section and meshes or openings of equal area, are secured to the guide tubes 5. Each of the fuel rods 7, 8, 9, 10 is guided through one respective mesh, and the grid-like spacers 6 combine the fuel rods into a bundle and hold them in such a way that they can freely expand axially. The laterally open construction of the fuel assembly 1 enables a crosswise mixing of the coolant K and makes for uniform heating of the coolant. FIG. 2 shows a complete cross section of a grid-like spacer 6 of a fuel assembly 1 of FIG. 1, as an example of a pressurized water reactor having 17.times.17=289 grid meshes. The fuel assembly 1 has 264 fuel rods with reference numerals 7, 8A and 10, twenty-four symmetrically distributed guide tubes 5, and one central instrumentation tube 5a. The drawing only shows fuel rods 7, 8A, 10 with the same mutual spacing P1 among the cross-sectional points, both in the longitudinal and the transverse direction of the grid (making a uniform distribution in a square basic configuration). For instance, the spacing P1 can be 12.6 mm. In other words, the spacing is basic to the grid, throughout the grid, and can be considered a standard spacing. For the sake of particularly good economical utilization of the nuclear fuel (for instance, uranium oxide, or in particular uranium plutonium mixed oxide), at least one fuel rod 8A (shown with shading inclined from the lower left to the upper right) in the immediate vicinity of a guide tube 5 and/or of the instrumentation tube 5A, should have a cladding tube with an outside diameter of 9.5 mm or 9.7 mm, for instance, which is larger than that of a fuel rod located farther away from the guide and instrumentation tubes 5, 5a. The thinner fuel rods which are located farther away from the guide and instrumentation tubes 5, 5a (with the shading inclined from the lower right to the upper left), are predominantly identified by reference numeral 7. In FIG. 5, the preferred case is shown, where all of the immediately adjacent fuel rods have the larger diameter, while all of the edge or peripheral meshes containing the edge fuel rods 10 and all of the other meshes adjoining the edge meshes contain thinner fuel rods. The group identified as "fuel rods located farther away" includes predominantly the four rows of edge fuel rods 10. The outside diameter of the cladding tubes of the fuel rods 7 located farther away is 9.14 or 9.30 mm, for example. In this exemplary embodiment, all of the fuel rods 7 located farther away, or in other words all those with shading oriented from the lower right to the upper left have a smaller diameter than the fuel rods 8A immediately adjacent the tubes 5, 5a. The group of "fuel rods 7 located farther away" can also predominantly (fully or partly) include the four adjacent fuel rods 7 located diagonally away. This depends on the desired uniformity. Accordingly, this exemplary embodiment also shows that the fuel rods 10 at the edge of the fuel assembly 1 have an outside diameter of the cladding tube of 9.14 or 9.30 mm, for instance, which is also smaller than all of the fuel rods 8A in the center of the fuel assembly 1. This center can be defined by the (almost circular) circumferential line described by the outer guide tubes 5. In FIG. 3, a portion of a fuel assembly cross section as seen from above is shown, in accordance with the first exemplary embodiment of FIG. 2. The fragmentary or detail view again shows the region in the fuel assembly 1 around a guide tube (or instrumentation tube 5a), but on a larger scale. It can be seen clearly in this case that the two fuel rods 8A shown immediately adjacent the guide tube 5 (or instrumentation tube 5a), have a larger outside diameter of the cladding tube than the fuel rods 7 located farther away from the tube 5 (or 5a). In a departure from the drawing, the fuel rod 7 located diagonally to the right from the tube 5 could also have a larger outside diameter of its cladding tube. Correspondingly, the two non-illustrated fuel rods 8A located to the left of and below the guide tube 5 could have a larger diameter than the fuel rods 7. As a result of this dimensioning of the four fuel rods 8A, more fuel can be accommodated in the fuel rods 8A near the tube 5, given an increased pellet diameter and the same wall thickness, for example, than in the fuel rod 7. As a result, the previously excessive local moderation in the region around the tube 5 (or 5a) is reduced. The dimensioning of the fuel rods 8A assures the uniform utilization of fuel over the cross section. The adjacent fuel rods 7 and/or 8A, located in a crosswise or lengthwise row, have spacings P1 of the cross-sectional centers or central axes thereof which are equal to one another. In other words, the center points or central axes of the fuel rods 7, 8A of each vertical or horizontal row are located along a straight line. Button and spring combinations for retaining the fuel rods 7, 8A in the various meshes are shown in this case at reference numeral 13. FIG. 4 shows a portion of a second embodiment of a fuel assembly cross section as seen from above. The second embodiment of the fuel assembly 1 is only insignificantly more difficult to manufacture than the embodiment of FIG. 3. Once again, particularly good utilization of the nuclear fuel is attained. The fragmentary or detail view shows an enlargement in the region of the fuel assembly 1 around a guide tube 5 (or instrumentation tube 5a). It can be seen that all of the fuel rods 7, 8B in this case are selected to have the same diameter. They preferably contain the same quantity of nuclear fuel, or in other words the same number of pellets of the same size. However, different weights are once again possible. It can also be seen that in this case the spacing P2 between a cross-sectional center point or point along the central axis M8 of the two fuel rods 8B that are immediately adjacent to and in the same row direction as the guide tube 5 (or instrumentation tube 5b) on one hand, and the cross-sectional center point M5 of the guide tube 5 (or instrumentation tube 5a) on the other hand is less than the standard spacing P1 of the cross-sectional center points M7 of (adjacent) fuel rods 7 located farther away. In other words P1 is larger than P2. The fuel rods 8B in this case are accordingly shifted closer to the tube 5, 5a. Accordingly, in the vicinity of the tubes 5, 5a a "non-uniform" distribution is obtained. In other words, the centers of the fuel rods and the guide rods describe the intersection points of a grid that is not uniform like the grid of the spacer but rather is deformed in the vicinity of the tubes 5, 5a, in order to assure that the moderation is made uniform in this case. The spacings P1, P2 are once again measured in the row direction. It should also be noted that the fuel rod 7 located in the diagonal direction can also be shifted closer to the tubes 5, 5a. This is also true for the three other non-illustrated adjacent fuel rods 7 in the diagonal direction. FIG. 5 shows a portion of a third embodiment of the corner of a fuel assembly cross section as seen from above. In this case a spacing P3 between the cross-sectional centers M9 of some fuel rods 9C in the second row and the cross-sectional center M10 of the adjacent fuel rod 10 at the edge of the fuel assembly (that is, on the edge or edge leg 11 of the spacer) is greater than the spacing P1, P2 of the inner fuel rods 7, 9C. That is, P3>P1>P2, and P3=2P1-P2. The fuel rods 7, 9C, 10 all have the same outside diameter. In other words, in FIG. 5 the fuel rods 9C of the second row in each case are shifted away from the adjacent edge fuel rod 10. Although it is not illustrated in FIG. 5, the fuel rod 7 in the second row can be shifted away from one or both adjacent fuel rods 10 of the edge row. In other words, the cross-sectional center M7 of the fuel rod 7 (shifted diagonally inward) can be located at the point P7. The diameters of all of the fuel rods 7, 9C, 10 are equal and the fuel rods can be standard fuel rods. Due to this construction, uniformity of the hot spots and a local increase in fuel utilization are also assured in the vicinity of the edges 11. In FIGS. 4 and 5, the "unshifted" fuel rods 7 or the fuel rods 7 and 10 form a first group, and the "shifted" fuel rods 8B or 9C form a second group. FIG. 6 shows how this result can be assured in a different way as well. In this case, each of the fuel rods 10 of the edge row has a smaller outside diameter than a fuel rod 7 located farther inward, which for instance is a standard fuel rod. Thus with the same cladding tube wall thickness, they contain less fuel. Moreover, the fuel rods 9D of the second row are again thinner than the fuel rods 7, which as noted above can, for instance, be standard fuel rods. Differing from the drawing, in this case, the fuel rods 9D and 10 can preferentially have the same outside diameter in order to reduce the number of types of fuel rods and thus keep manufacturing costs low. The spacings P1 between all of the adjacent fuel rods 7, 9D and 10 are equal. The effect of the construction chosen in FIG. 6 is equivalent to that of the construction chosen for FIG. 5, if the spacings P2, P3 are suitably selected in FIG. 5. FIG. 7 shows two adjacent fuel rods 12a and 12b having cladding tubes with the same outside diameter D1, in which one fuel rod 12a has a different ratio s1/D1 of its cladding tube wall thickness s1 to its cladding tube outside diameter D1 than the adjacent fuel rod 12b. The ratio s2/D2 is larger than s1/D1. The fuel rod 12a can therefore hold more fuel than the fuel rod 12b. FIG. 8 shows two adjacent fuel rods 12c, 12d with cladding tubes having different outside diameters D2 and D3 (D3 is larger than D2). It is assumed in this case that the wall thickness s3 is the same for the cladding tubes of both fuel rods 12c, 12d. Accordingly the following applies to the ratios: s3/D2>s3/D3. In this case, the fuel rod 12d can hold more fuel than the fuel rod 12c. FIG. 9 shows two adjacent fuel rods 12e and 12f, where the relationship between the outside diameters D4 and D5 of their cladding tubes are such that D4 is smaller than D5, while the relationship of their cladding tube wall thicknesses s4 and s5 are such that s4<s5. The dimensioning is selected in such a way that the following applies: s4/D4=s5/D5. This assures mechanical stability. In this case, the fuel rod 12e can contain more fuel than the fuel rod 12f. As shown in FIG. 10, a pressurized water reactor has a pressure vessel 22, in which a reactor core with diagrammatically illustrated nuclear reactor fuel assemblies 1 is located. The fuel assemblies have fuel rods in accordance with an exemplary embodiment as described above. An outlet 24 for water (both coolant and moderator) from the pressure vessel 22 and thus from the reactor core having the nuclear reactor fuel assemblies 1, is connected to one end of a primary tube 26 of a steam generator 27. An inlet 25 for leading the water into the pressure vessel 22 and thus into the reactor core is connected to the other end of the primary tube 26. The primary loop formed by the pressure vessel 22 and the primary tube 26 is a closed loop, so that no steam can form in this primary loop and thus in the reactor core. Steam does form on the secondary side of the steam generator 27, which has a delivery connection 28 for feedwater 29 and an outlet connection 30 for steam. The steam may, for instance, be carried from the outlet connection 30 to a non-illustrated steam turbine.
	summary
	summary
048448575	summary	The invention relates to a pressurized water reactor and, more particularly, to a reactor with a primary circuit including therein a reactor pressure vessel, a steam generator and a main coolant pump, and with an auxiliary system having high-pressure pumps for feeding water into the primary circuit, the auxiliary system being connected with the reactor pressure vessel through a line extending from the upper side of the latter and having at least one shut-off or isolating valve therein. In installations of the afore-described type, cold water was heretofore introduced into the reactor pressure vessel, to condense steam which may be present therein as a result of a leak. This requires pumps with a feed pressure higher than the steam pressure in the reactor pressure vessel. The cold water serving to effect the condensation can lead to thermal stresses, however, which are all the more undesirable whenever there is a failure. It is accordingly an object of the invention to provide a pressurized water reactor which, in case of a power failure, when the main cooling pumps do not operate, and cooling of the reactor pressure vessel has broken down, removes hot cooling liquid in the upper part of the reactor pressure vessel without introducing cold water into the reactor pressure vessel by means of strong pumps. With the foregoing and other objects in view, the invention which utilizes a technique exactly opposite that used in the heretofore known state of the art provides for a pressurized water reactor with a primary circuit including therein a reactor pressure vessel, a steam generator and a main coolant pump, and with an auxiliary system having high pressure pumps for feeding water into the primary circuit, comprising a line extending from an upper side of the pressure vessel and having at least one shut-off valve therein, the line connecting the reactor pressure vessel and a part of the auxiliary system wherein a lower pressure prevails than in the reactor pressure vessel. With such construction, an outlet for the hot coolant is created, and the hot coolant is cooled outside of the reactor pressure vessel. In accordance with another feature of the invention, the line is connected to the suction side of a water-jet pump, which is supplied and operated with the water of the auxiliary system as the working or driving fluid. The water-jet pump effects an intimate mixing with the colder water of the auxiliary system. Furthermore, a defined negative pressure can thereby be generated. In accordance with a further feature of the invention, the auxiliary system used a volume control system of the pressurized water reactor, because the volume control system already had adequately dimensioned pumps for feeding water into the primary circuit. With these pumps, the coolant can be returned to the primary circuit, so that the liquid level therein is not reduced. In accordance with an alternate feature of the invention, the auxiliary system is a boron-addition system of the pressurized water reactor, which is likewise provided with high pressure pumps. Furthermore, the auxiliary systems, regardless of the construction thereof, can be connected with the reactor pressure vessel by a motor-driven pump, for example, a centrifugal pump, in order to remove coolant by suction in a metered or dosed manner. In accordance with an added feature of the invention, a temperature measuring station is provided in the line, so that the temperature of the discharging coolant can be used for controlling the quantitative discharge flow hereof, or the like. In accordance with a concomitant feature of the invention, the pressure vessel has a conventional cover closing the top thereof, and the line extends from the cover of the pressure vessel. 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 pressurized water reactor, 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 or 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, in which:
056549920	abstract	A neutron-irradiated structural material having a defect in a nuclear reactor internals is repaired by covering the structural material to be repaired with a plate to cover over a portion thereof having the defect, and then welding the plate and the structural material by locally applying pressure on the surface of the plate and adding energy to the portion to which the pressure is applied thereby to generate thermal energy in the contact surfaces between the plate and the structural material.
	claims	1. A high-temperature gas reactor control rod comprising:first and second control rod elements each having a neutron absorber between an outer cylinder and an inner cylinder that form a double cylindrical tubular shape along a longitudinal direction of each control rod element;first and second columnar support members respectively disposed in the inner cylinder of the first and second control rod elements, each of the first and second columnar support members axially extending in each control rod element in its entire length of the longitudinal direction to allow the first and second columnar support members to be vertically aligned and connected, thereby joining the first and second control rod elements in the longitudinal direction; anda joint structure provided to at least one of the first and second columnar support members and configured to connect a lower end of the first columnar support member and an upper end of the second columnar support member,wherein each of the first and second control rod elements has a neutron absorber supporting flange extending toward the inner cylinder, the neutron absorber supporting flange being formed on each support member, each support member supporting the load of the neutron absorber from below by the neutron absorber supporting flange,each support member has a side wall assembly and a hollow portion, the side wall assembly having a plurality of flat plates which define and form the hollow portion inside each support member, the hollow portion extending in the longitudinal direction,the plurality of flat plates are made of composite materials which are selected from the group consisting of C/C composite materials and SiC/SiC composite materials, andthe joint structure comprises a shaft and a loop-shaped joining belt, the shaft being inserted through the loop-shaped joining belt and two holes formed in the side wall assembly in each of the first and second support members to connect the lower end of the first columnar support member and the upper end of the second columnar support member. 2. The high-temperature gas reactor control rod according to claim 1, wherein the joining belt forms a flat belt shape. 3. The high-temperature gas reactor control rod according to claim 1, wherein a gap is provided between the control rod elements. 4. The high-temperature gas reactor control rod according to claim 1, wherein the joint structure is provided respectively at both upper and lower ends of the first or second support member, the shaft provided at the upper end and the shaft provided at the lower end are disposed so as to extend in different directions in a planar view, and the width of the joining belt is configured to be smaller than the axial length of the shaft in the hollow portion. 5. The high-temperature gas reactor control rod according to claim 1, wherein the diameter of the shaft is smaller than the diameter of the two holes formed in the side wall assembly so as to form a clearance for alleviating stress. 6. The high-temperature gas reactor control rod according to claim 1, wherein an internal hollow portion width of the joining belt along a direction perpendicular to the axis of the shaft is larger than the diameter of the shaft so as to form a clearance for alleviating stress. 7. The high-temperature gas reactor control rod according to claim 1, wherein the joint structure comprises a first shaft, a second shaft, and a loop-shaped joining belt, the first shaft is inserted through the loop-shaped joining belt and holes formed at a position near the lower end of the side wall assembly of the first support member in the first control rod element, and the second shaft is inserted through the loop-shaped joining belt and holes formed at a position near the upper end of the side wall assembly of the second support member in the second control rod element. 8. The high-temperature gas reactor control rod according to claim 1, wherein a length of the shaft (E), a distance between outer ends of the two holes (D), and a space between each outer end of the two holes and an inner surface of the inner cylinder (S) meet the following relationship (1):S+D≦E≦2S+D (1). 9. The high-temperature gas reactor control rod according to claim 1, further comprising a neutron absorber supporting flange extending toward the inner cylinder, the neutron absorber supporting flange formed at the lower end of the first support member or in the vicinity thereof. 10. The high-temperature gas reactor control rod according to claim 1, wherein the support members have a polygonal cross-sectional shape, and the polygon is an even number polygon. 11. The high-temperature gas reactor control rod according to claim 10, wherein the support members in the polygonal cross-sectional shape are formed by combining flat plate-shaped parts made of a carbonaceous material with each other. 12. The high-temperature gas reactor control rod according to claim 11, wherein the neutron absorber supporting flange is integrally formed with the flat plate-shaped part made of a carbonaceous material by forming the neutron absorber supporting flange together with the flat plate-shaped part as one solid piece, so as to protrude in the shape of a rectangular piece. 13. The high-temperature gas reactor control rod according to claim 1, wherein the neutron absorber supporting flange is integrally formed with the flat plate-shaped part made of either one of a C/C composite material and a SiC/SiC composite material by forming the neutron absorber supporting flange together with the flat plate-shaped part as one solid piece, so as to protrude in the shape of a rectangular piece. 14. The high-temperature gas reactor control rod according to claim 1, wherein the joint structure is configured to join the first and second control rod elements by bolt connection. 15. The high-temperature gas reactor control rod according to claim 11, wherein the neutron absorber supporting flange is integrally formed with the flat plate-shaped part made of a carbonaceous material by forming the neutron absorber supporting flange separately from the flat plate-shaped part as different pieces to join the neutron absorber supporting flange to the flat plate-shaped part, so as to protrude in a shape of a rectangular piece. 16. The high-temperature gas reactor control rod according to claim 1, wherein the neutron absorber supporting flange is integrally formed with the flat plate-shaped part made of either one of a C/C composite material and a SiC/SiC composite material by forming the neutron absorber supporting flange separately from the flat plate-shaped part as different pieces to join the neutron absorber supporting flange to the flat plate-shaped part, so as to protrude in a shape of a rectangular piece.

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