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050230433	claims	1. An actively cooled heat shield for use at a high temperature range, as is present in a fusion reactor, comprising: at least one body made of a refractory material, said body having a thermally stressed surface facing the fusion reactor and an opposing surface, the opposing surface having at least one recess, the at least one recess being at least partially arc-shaped in cross-section: a respective metal cooling pipe disposed in the respective at least one recess and thermally coupled thereto, the cooling pipe having a cross-section complementary to the cross-section of the respective recess in which the cooling pipe is disposed, forming a heat transfer surface, the cooling pipe being brazed along the entire heat transfer surface to the refractory body; and a cooling medium circulating in the cooling pipe. 2. The heat shield device as defined in claim 1, including a plurality of said bodies which are closely aligned and each of which have at least one cooling pipe brazed thereto. 3. Devices according to claim 2, wherein at least one refractory body has a plurality of said recesses which are of different sizes and in which a respective plurality of cooling pipes having different diameters are brazed. 4. Device according to claim 2, wherein the bodies have elongated shapes, whereby conductive loops cannot be established in electromagnetic fields present in the fusion reactor. 5. Device according to claim 1, wherein the recess is a groove having an at least partial cylindrical surface and the complementary cooling pipe has a circular cross section. 6. Device according to claim 5, wherein the recess has a U-shaped cross section and is deeper than one-half the exterior diameter of the complementary cooling pipe. 7. Device according to claim 2, wherein the bodies each have at least one recess configured as a bore into which a cooling pipe is brazed. 8. Device according to claim 1, wherein the refractory material is selected from the group consisting of graphite, carbides, ceramics, and composite metal-ceramic materials. 9. Device according to claim 1, wherein cooling pipes brazed to the at least one body are each composed of a metal having essentially the same coefficient of thermal expansion as the refractory material and are each connected to a coolant conduit made of another material. 10. Device according to claim 9, wherein the connection between each cooling pipe and the coolant conduit includes a shrink fitted seat around the cooling pipe. 11. Device according to claim 1, wherein each refractory body is supported mechanically by the cooling pipes to which it is brazed. 12. Devices according to claim 1, wherein the at least one refractory body has a plurality of said recesses which are of different sizes and in which a respective plurality of cooling pipes having different diameters are brazed. 13. Device according to claim 1 wherein the at least one body has the shape of an elongated strip. 14. Device according to claim 2, wherein the bodies are juxtaposed to one another with interstices therebetween, said interstices having such a configuration that thermal radiation from a heat source to be shielded is unable to pass in a straight line through the interstices. 15. Device according to claim 8, wherein cooling pipes brazed to the respective at least one body are each composed of a metal having essentially the same coefficient of thermal expansion as the refractory material and are each connected with a coolant conduit made of another material. 16. Device according to claim 1 wherein the refractory material is selected from the group consisting of graphite, a carbide, and beryllium. 17. Device according to claim 1, wherein the refractory material is selected from the group consisting of graphite, SiC, TiC, B.sub.4 C, and beryllium. 18. Device according to claim 1, wherein each respective body of refractory material has the form of a solid block.
	abstract	A system of shields designed to provide substantially greater protection, head to toe, against radiation exposure to health care workers in a hospital room during procedures which require real-time imaging. The shields are placed around the patient and the x-ray table and provide protection even when the x-ray tube is moved to various angles around the patient.
	abstract	A method of using a radiation protection system including a table having a top surface for supporting a patient, a radiation-shielding screen attached to the table for covering a portion of the patient and a portion of the top surface of the table, and controls for controlling the system. The radiation-shielding screen includes at least one port. The method includes extending the radiation-shielding screen over a portion of the patient the table and accessing the controls through the port. The method further includes controlling the system using the controls.
	abstract	The invention comprises a method and apparatus for using a multi-layer multi-color scintillation based detector element to image a tumor of a patient using a process of determining residual energies of positively charged particles after passing through the patient, the process comprising the steps of: (1) transmitting the positively charged particles at known energies through the patient and into a multi-layer detector element; (2) detecting first and second secondary photons, resultant from passage of the positively charged particles, respectively from a first layer of a first scintillation material and a second layer of a second scintillation material at two respective layer depths, where the first wavelength range differs from the second wavelength range; (4) determining residual energies of the positively charged particles, using output from the step of detecting; and (5) relating the residual energies to body densities to generate an image.
	description	The present invention relates to a process for immobilizing metallic sodium in glass form, that can be used in particular for the vitrification of metallic sodium containing radioactive elements, such as the sodium employed in the liquid state in certain nuclear plants, like for example in the primary or secondary cooling systems of nuclear reactors of the fast-neutron type. Liquid sodium that has been employed in a nuclear plant may have variable contents of radioactive elements, such as 22Na, tritium or 137Cs. Thus, the sodium present in certain test loops of pilot reactors may contain more than 0.001 wt % of radioactive elements, whereas sodium that has been employed in a primary or secondary cooling system of a conventional nuclear reactor of the fast-neutron type is generally a slightly radioactive sodium, usually containing of the order of a few 10−12 g of radioactive elements per gram, these radioactive elements furthermore having a short half-life and stable daughter products. In any case, whatever its content of radioactive elements, sodium that has been employed in a nuclear reactor is considered, as regards the current legislation, in such form as a nuclear waste that may potentially contaminate the environment. Thus, even when it has an extremely small content of radioactive elements, this sodium is at least classed among nuclear waste referred to as being “of very low activity” (VLA). Consequently, during operations to dismantle nuclear plants employing sodium, it is necessary to store this sodium so as to prevent it from coming into contact with the external environment, whatever its degree of contamination. In this context, it has been proposed to immobilize the sodium in a salt, silicate or oxide form within a matrix, such as a cement or a glass, in a similar manner to the storage of many radioactive or toxic wastes. However, owing to the high solubility of sodium compounds in water, one specific problem that arises when immobilizing sodium within a matrix of glass or cement type is that of the potential leaching of the sodium, namely the tendency that a glass or cement matrix may have of releasing the sodium compounds in the presence of water. Under the current standards regarding the storage of radioactive waste, it is necessary, in order to immobilize the sodium, to move towards matrices that have the lowest possible tendency of sodium leaching. From this standpoint, glass-type matrices are generally the most promising. In the present description, the term “glass” should be taken in its most widely accepted meaning within the field of waste vitrification, that is to say a relatively homogeneous solid mixture of various mineral constituents, generally based on silica, in the form of a non-porous solid, advantageously formed from chemically and structurally bonded elements, and suitable for waste immobilization. The term “glass” within the meaning of the invention preferably denotes a solid having specifically an amorphous structure. In certain circumstances, this term may nevertheless also denote solids having a partially crystalline structure, especially of the vitreous-crystalline type, or even crystalline solids. When employed within the meaning of the present description, the term “glass-type matrix” denotes any solid matrix having a continuous phase based on a “glass” within the meaning of the invention. Glass-type matrices have a much lower specific surface area than cement-type matrices, especially in so far as they are not porous. Consequently, they provide a greatly reduced area for exchange with the external medium. However, it should be emphasized that, in order for a glass-type matrix containing sodium compounds to be really promising in terms of limiting sodium leaching, it is generally desirable for the sodium compounds that it contains to be incorporated in the form of effective constituents of the matrix, and not as inclusions, otherwise the sodium compound is merely physically encapsulated within the glass-type matrix. In this case, the sodium compound is relatively weakly integrated into the matrix, thereby resulting in a very marked increase in the tendency of this compound to leach out. More generally, to avoid the phenomenon of sodium within a glass-type matrix leaching out, it is desirable for the composition of this sodium-based matrix to be as homogeneous as possible. Thus, to obtain really promising vitrified matrices, it has therefore proved necessary to immobilize the sodium within these matrices under the most controlled conditions possible. Most of the suggested solutions for achieving such immobilization of the sodium under controlled conditions consist in chemically converting the sodium into derivatives that can be incorporated into a vitrified matrix, in particular of the sodium carbonate or sodium oxide type, then in introducing these sodium derivatives into a glass formulation, before carrying out the vitrification of the mixture obtained under controlled conditions. However, the operations of chemically converting the sodium that are used in this context lead to the formation of liquid and/or gaseous effluents liable to result in radioactive elements being released into the environment. In addition, most of the radioactive sodium treatment operations proposed have the drawback of involving several successive steps, which, in addition to requiring the intermediate storage of radioactive compounds, further increases the number of liquid or gaseous effluents to be controlled, this being manifested not only in terms of high production costs, but above all in terms of increasing the risk of uncontrolled discharge into the environment. So as to avoid such problems associated with increasing the number of liquid or gaseous effluents, one promising solution, described for example by R. Kushar et al. in the report ANL-91-21 published in 1991 by the Argonne National Laboratory, consists in mixing, within a “cyclone” reactor heated to a temperature of 1000° C., silica, alumina, quick lime and boron oxide particles with metallic sodium introduced in the form of a dispersion of droplets in the liquid state, in the presence of a stream of air introduced in an amount sufficient to oxidize the metallic sodium into the form of sodium oxide. This process makes it possible to obtain vitrified matrices of homogeneous composition and able to have low leaching rates depending on the amounts of silica, alumina, quick lime and boron oxide introduced. However, this process involves the use of large quantities of air, only some of which is consumed by the sodium oxidation reaction. This has the drawback of resulting in the production of large quantities of gaseous effluents from the reactor, which effluents are likely to convey radioactive dust, which, here again, is reflected in terms of effluent treatment costs and potential repercussions for the environment. More generally, at the present time a vitrification process without appreciable production of gaseous elements, making it possible within a single reactor to prepare, from metallic sodium, a vitrified matrix of homogeneous composition incorporating the sodium in the form of an oxide, is unknown. Now, the inventors have discovered that such a process could be realized by introducing, into a vitrification reactor, metallic sodium, a precursor of the final vitrified matrix, and an iron oxide Fe2O3 as oxidizing agent for the sodium provided that the sodium, the iron oxide and the precursor of the mineral matrix are specifically introduced into the reactor in a dispersed form. The inventors have demonstrated that the use of the various constituents in the dispersed state makes it possible not only to optimize the reaction of sodium oxidation by iron oxide Fe2O3, but also to obtain in fine a vitrified matrix of homogeneous composition. On the basis of this discovery, it is an object of the present invention to provide a process for immobilizing sodium within a vitrified matrix, this process being suitable for the containment of sodium containing radioactive elements. Within this context, one of the objects of the invention is in particular to provide a sodium immobilization process that limits any possible exchange between the contaminated sodium and the external medium. More generally, it is also an object of the invention to provide a process for immobilizing metallic sodium in the form of a glass, this process being both simple to implement and one in which the final composition of the vitrified matrix obtained can be easily varied according to the desired properties of this matrix, especially in terms of resistance both to leaching and to devitrification. Thus, the subject of the present invention is a process for immobilizing metallic sodium in glass form, comprising the steps consisting in: (A) introducing, into a reactor: (i) a vitrified matrix precursor, (ii) sodium in the metallic state, and (iii) iron oxide Fe2O3, in an amount sufficient to ensure oxidation of the metallic sodium (ii) introduced, the said constituents (i), (ii) and (iii) being especially introduced into the reactor in a dispersed form; (B) producing, by bringing constituents (i), (ii) and (iii) in dispersed form into contact with each other within the reactor, a homogeneous mixture of these constituents; (C) heating the mixture obtained to a temperature of between 1000 and 1600° C., preferably greater than 1200° C. and advantageously greater than 1400° C., whereby a homogeneous mixture in the molten state is formed, in which the sodium initially introduced is in the sodium oxide state; and (D) recovering the molten mixture thus obtained and cooling it, whereby a vitrified matrix of homogeneous composition is obtained that incorporates the initially introduced sodium as constituent oxide of the said vitrified matrix. Within the meaning of the invention, the term “vitrified matrix precursor” is understood broadly to mean an assembly of constituents, generally solids, that are preferably inert at low temperature with respect to sodium and iron oxide, and are capable of resulting in the formation of a vitrified matrix after the heat treatment step (C) and after the cooling of step (D), in the presence of the products resulting from the reaction of sodium with iron oxide Fe2O3. In the most general case, the mineral matrix precursor (i) of the invention may consist of any mixture of particles based on oxides, carbonates or salts commonly used in the manufacture of glasses. In this context, it may in particular be chosen from the mixtures generally denoted by the term “batch materials” employed in the manufacture of “nuclear” glasses used for the containment of radioactive wastes. Thus, the vitrified matrix precursor (i) may, for example, be formed at least partly from solid particles comprising particles based on silicon oxide SiO2, generally in combination with particles based on calcium oxide CaO and/or boron oxide B2O3, optionally in combination with particles based on aluminium oxide and/or magnesium oxide, the particles based on the various aforementioned oxides possibly being particles based on only one of these oxides, or based on a combination of two or more of these oxides. Especially so that the vitrified matrix obtained by the process of the invention has the least possible tendency to leaching and devitrification, it is generally preferred for the vitrified matrix precursor of the invention to include a mixture of particles based on: silicon oxide SiO2; and calcium oxide CaO and/or boron oxide B2O3, with SiO2/(CaO+B2O3) mass ratios generally between 1 and 10, and preferably between 5 and 8. Boron oxide B2O3 may in particular be used to reduce the melting point of the vitrified matrix precursor (i) or to improve the mechanical integrity of the matrix during the cooling of step (D). In particular so as to further improve the properties of the glass matrix finally obtained, the precursor (i) may advantageously furthermore include compounds chosen from aluminium oxide Al2O3, magnesium oxide and zinc oxide. Preferably, the vitrified matrix precursor of the invention comprises: silicon oxide in an amount from 20 to 80 wt % and preferably between 40 and 75 wt %; calcium oxide in an amount from 0 to 20 wt % and preferably between 1 and 10 wt %; alumina in an amount from 0 to 20 wt % and preferably between 1 and 10 wt %; boron oxide B2O3 in an amount from 0 to 20 wt % and preferably between 1 and 10 wt %. Whatever its precise composition, the vitrified matrix precursor (i) is specifically introduced in a dispersed form into the reactor where the process of the invention takes place. Thus, it is generally introduced in the form of a mixture of particles, generally metal oxide particles, or particles of preformed glass frits, these particles preferably having a mean size of between 0.1 and 20 millimeters, advantageously less than 10 millimeters and preferably less than 5 millimeters. Thus, the vitrified matrix precursor of the invention may advantageously be employed in the form of a mixture of silica, alumina, quick lime and boron oxide particles having particle sizes of between 1 and 3 millimeters. It may also be employed in the form of glass frits, especially in the form of glass frits based on silicon, aluminium, calcium and boron oxides, these glass frits advantageously having a mean size of between 1 and 10 millimeters, and preferably less than 5 millimeters. It should be emphasized that, given the fact that the process of the invention does not lead to an appreciable quantity of effluents being produced, this process may be carried out under rigorous containment conditions, particularly those suitable when handling radioactive compounds. Consequently, the metallic sodium used in the process of the invention may advantageously be a sodium containing radioactive elements, and it may especially be a sodium coming from a primary or secondary cooling system of a nuclear reactor of the fast-neutron type. Whatever its exact nature, the sodium used according to the process of the invention is specifically introduced into the reactor in a dispersed form. Thus, according to one particularly advantageous embodiment, the sodium may be introduced into the reactor in the form of liquid droplets, preferably having a size of between 10 microns and 500 microns, advantageously less than 200 microns and preferably between 50 and 150 microns, these droplets usually being obtained by spraying the liquid sodium, for example by means of an injection nozzle. According to another possible embodiment, the sodium introduced may also come from premixing the liquid sodium in the dispersed state with at least part of the vitrified matrix precursor. In this case, the sodium in the dispersed state that is introduced is in the form of sodium deposited on vitrified matrix precursor particles. Thus, according to this particular method of implementation, the sodium in the dispersed state that is introduced in step (A) of the process may, for example, be in the form of silica particles on the surface of which sodium has been deposited. In general, especially so as to obtain a vitrified matrix having the lowest possible tendency to leaching after the process of the invention, it is often preferable that the amount of sodium incorporated into this matrix be less than 30% by mass, and advantageously less than 20% by mass, with respect to the total mass of the matrix. It is also preferable that this amount of sodium incorporated be greater than 4% by mass, and advantageously more than 5% by mass, with respect to the total mass of the matrix. Thus, this amount is typically between 5% and 15% by mass with respect to the total mass of the matrix. Consequently, it is generally preferable for the mass of metallic sodium introduced into the reactor to be between 3% and 20% by mass with respect to the mass of vitrified matrix precursor introduced. Advantageously, this mass of sodium introduced is less than or equal to 14% by mass, and preferably less than or equal to 12% by mass, with respect to the mass of vitrified matrix precursor introduced. Moreover, in the particular case in which the vitrified matrix precursor is based on silicon oxide, which is generally the case, it is preferable for the molar ratio of the amount of sodium introduced to the amount of silicon introduced into the vitrified matrix precursor to be between 1/20 and 1/1, and advantageously less than 1/6. One of the essential constituents used in the process of the invention is iron oxide Fe2O3. This constituent ensures that, within the reactor, the sodium is converted to the oxide, in which form the sodium is effectively incorporated within the vitrified matrix, specifically as a constituent element of this matrix, and not as an inclusion element. Iron is also a network former, in the same way as silicon or aluminium. In particular, iron, in the presence of these elements silicon and aluminium, forms the glassy network by interconnections. It can also result in the formation of mineral structures known to include cavities that can trap certain radioactive contaminants possibly present in the sodium. It is highly preferable for the iron oxide Fe2O3 to be introduced into the reactor in a dispersed form, especially so as to maximize the number of interfaces for contact between this oxide and the sodium, which is also introduced in a dispersed form offering a high exchange surface area. As a result, it is generally preferable for the iron oxide Fe2O3 to be introduced in the form of particles comprising iron oxide Fe2O3, these particles preferably being essentially formed from iron oxide Fe2O3, and advantageously having a mean size of between 0.1 and 20 millimeters, particularly preferably less than 10 millimeters, and advantageously less than 5 millimeters. It is preferable for the specific surface area of these particles to be as high as possible. The amount of iron oxide Fe2O3 introduced is also a key parameter in order to ensure effective oxidation of the sodium introduced in the metallic state. This is because it is necessary for the iron oxide Fe2O3 to be introduced in an amount sufficient to ensure that this oxidation takes place. In this context, it is generally preferable for the molar ratio of the amount of iron introduced in the form of iron oxide Fe2O3 to the amount of sodium introduced to be between 0.5/1 and 3/1. Preferably, this molar ratio is greater than or equal to 0.6/1, and advantageously greater than or equal to 0.9/1. In general, it is also preferable for this molar ratio to be less than 2/1 and advantageously less than 1.5/1. Without wishing to be tied in any way to one particular theory, it seems possible to suggest that the sodium oxidation reaction that takes place within the reactor involves the following reactions:2Na+Fe2O3→Na2O+2FeO2Na+3Fe2O3→Na2O+2Fe3O4 In any case, the vitrified matrix obtained in fine generally contains iron oxides FeO and/or Fe3O4 or even residual oxides Fe2O3. In particular when the process of the invention is carried out for the containment of sodium that includes radioactive elements, it is usually necessary to control the amount of iron introduced into the vitrified matrix produced according to the invention. For this purpose, it is generally preferable for the mass of Fe2O3 introduced to be between 5% and 50% by mass with respect to the mass of vitrified matrix precursor introduced, and preferably in an amount of less than 35% by mass, advantageously less than 30% by mass, and more preferably less than 25% by mass. Whatever their exact nature and their respective proportions, constituents (i), (ii) and (iii) of step (A) of the process of the invention are generally introduced without any trace of water, so as to avoid any reaction between sodium and water, which would be liable to result both in the production of a hydrogenated gaseous effluent and in embrittlement of the vitrified matrix finally obtained. To do this, prior to their use in the process, the vitrified matrix precursor and the iron oxide Fe2O3 employed are generally dried and then stored away from any trace of moisture. The solids introduced into the reactor may be conveyed by small amounts of carrier gas. In this case, the carrier gas is generally free of any trace of water and is preferably an inert gas, such as nitrogen. This carrier gas is likely to lead to the formation of only small volumes of gaseous effluents, this being particularly advantageous when the process is carried out on sodium containing radioactive elements (limitation of secondary effluents). Moreover, constituents (i), (ii) and (iii) of step (A) of the process of the invention are generally introduced into the reactor at a temperature ranging from 15° C. to 150° C., the sodium generally being introduced in the liquid state, that is to say at a temperature preferably greater than 100° C., usually between 110° C. and 130° C., and the other constituents being able to be introduced at lower temperatures, for example between 15° C. and 100° C., typically between 20 and 60° C. The homogeneous mixture of step (B) of the process, produced by bringing the various constituents (i), (ii) and (iii) into contact with each other in dispersed form, is generally carried out by introducing the vitrified matrix precursor and the iron oxide into the reactor in the form of a rain of particles and by spraying the sodium in the liquid state into the said rain of particles. The term “rain of particles” is understood, according to the invention, as meaning a descending dispersion of particles, advantageously dispersed homogeneously, of the type of those used in reactors employing reactants in powder or dispersed form. Rains of particles according to the invention may be produced using any type of standard powder dispersion system known in the prior art. The rain of particles based on the vitrified matrix precursor and on iron oxide is advantageously produced from an initial mixture of iron oxide (iii) with the vitrified matrix precursor (i), but it may alternatively be obtained by interpenetration of two or more rains of particles based on iron oxide and vitrified matrix precursor. Another method of producing the mixture of step (B) consists in depositing the sodium on at least some of the vitrified matrix precursor particles, generally on silica particles, then in mixing constituents (i), (ii) and (iii) by interpenetration of two rains of particles comprising, in the case of one of them, particles based on Fe2O3, and, in the case of the other one, the vitrified matrix precursor particles, at least some of which have been impregnated with sodium. Whatever its exact method of implementation, it is generally preferable, especially so as to optimize the in situ sodium oxidation reaction, for the homogeneous mixture produced in step (B) to be heated to a temperature of between 150° C. and 400° C. prior to the heat treatment of step (C). Thus, it is generally preferable for constituents (i), (ii) and (iii) to be homogeneously mixed at a temperature below 150° C. and then to be brought under the temperature conditions of step (C), that is to say to a temperature of 1000–1600° C., preferably between 1400 and 1500° C., with an increasing temperature gradient. For this purpose, the reactor used in the process of the invention preferably has an upper part at a temperature of between 100° C. and 150° C., where mixing step (B) takes place, and a lower part under the temperature conditions of step (C), that is to say at a temperature of 1000–1600° C., and preferably at a temperature between 1400 and 1500° C., the constituents for step (A) being introduced into the upper part of the reactor where they are mixed. According to one particular method of implementation, the reactor used in the process of the invention comprises a feed zone at a temperature of 100° C. to 200° C., a sodium oxidation zone at a temperature of 150 to 400° C., and a melting zone under the temperature conditions of step (C), that is to say at a temperature of 1000–1600° C., and preferably 1400 to 1500° C. In this case, it may prove preferable in the sodium oxidation zone for the reactor to have a frustoconical-type cross section that flares out downwards or, more generally, in the said sodium oxidation zone, for the cross section of the reactor to increase on going from the top of the reactor downwards. This is because, in so far as the oxidation of sodium is an exothermic reaction, premature vitrification leading to congealing of the reactiron mixture may be observed in the oxidatiron zone. In this situatiron, the preferred configuratiron of the reactor proposed above allows the crust of glass formed to flow towards the lower zone of the reactor under the temperature conditirons of step (C), that is to say at a temperature of 1000–1600° C., where it is melted within the reactiron mixture. This prevents a blockage of the constituents (i), (ii) and (iii) at the top of the reactor, something which would have the effect of reducing the effectiveness of the mixing of these constituents and/or of limiting the efficiency of the sodium oxidatiron reactiron. Whatever the exact configuratiron of the reactor used in the process of the inventiron, the heat treatment step (C) may be carried out using any means, known to those skilled in the art, suitable for producing a vitrified matrix. The residence time of the mixture under the temperature conditions of step (C) must of course be long enough to bring the homogeneous mixture from step (B) under the said temperature conditions. Step (C) of the process of the invention may be carried out, for example, by heating the walls of the reactor within which the process of the invention is carried out, especially by resistance heating. However, it is particularly advantageous for step (C) of the process of the invention to be carried out by means of induction heating. The use of induction heating within the context of producing a glass is a known technique and one that has been widely described, this being based on the fact that the glass, which is insulating at low temperature, becomes conducting under the effect of an increase in temperature. The use of induction heating is particularly advantageous in so far as it can allow a layer of cooled glass to form on the walls of the reactor, this layer acting as a protective layer for the reactor throughout the duration of the process, this proving particularly beneficial if it is desired to carry out the process while ensuring optimum containment conditions. Such a technique, generally referred to as induction heating “in a cooled self-crucible”, is also a technique widely used within the context of the vitrification of radioactive waste, and especially waste of any activity. For more information on this subject, reference may in particular be made to the articles D 5935, D 5936 and D 5937 by G. Delevey, published in Les techniques de l'ingénieur [Engineering techniques]”, Génie électrique [Electrical Engineering], Volume D12. According to one particular method of implementing the process, it is possible, during step (C), to convert at least some of the iron oxides, FeO and/or Fe3O4, resulting from the sodium oxidation reaction into the form of the oxide Fe2O3, which is more effective in forming the glass matrix than the oxides FeO or Fe3O4. To do this, oxygen is injected in a controlled manner into the mixture during step (C). The amount of oxygen used in this context is, however, preferably adapted so that all of the oxygen introduced is consumed i.e. so that the oxygen introduced is capable neither of leading to the formation of gaseous effluents nor of resulting in bubbles within the final vitrified matrix. It is often preferable to use particles (iii) of iron oxide Fe2O3 having the highest possible specific surface area. Although steps (A) to (C) of the process of the invention are key in obtaining a homogeneous composition within the vitrified matrix obtained according to the process of the invention, step (D) of forming the vitrified matrix from the molten mixture resulting from step (C) is itself an important step, especially as regards the physical properties of the vitrified matrix obtained. In particular if the process of the invention is intended for the containment of sodium containing radioactive elements, it is necessary for the vitrified matrix obtained to have as small a surface area as possible for exchange with the external environment. For this purpose, it is therefore necessary for the vitrified material obtained to have a high tensile strength. Step (D) generally consists of an operation in which the mixture in the molten state (magma) from step (C) is cast into a mould or into metal containers (generally steel drums). According to another method of implementation, the magma obtained by casting may be rolled. The cooling of the composition generally takes place under conditions such that they limit the risk of microcracks appearing by thermal shock. In this context, it is generally preferable for the composition to be cooled as slowly as possible. For this purpose, it may be envisaged, if necessary, to reduce the temperature in successive stages. Another means of limiting the risk of microcracks appearing by thermal shock consists in using boron oxide B2O3 as a constituent of the mineral matrix precursor (i). According to one particularly advantageous method of implementation, the process of the invention may be carried out continuously, that is to say by a continuous addition of constituents (i), (ii) and (iii) into the reactor, and continuous casting of the mixture resulting from step (C) after the reactor. This possibility of carrying out the process continuously constitutes a certain advantage within the specific context of the vitrification of metallic sodium containing radioactive elements. This is because it allows the sodium to be treated using a completely confined process. Within the context of a process carried out continuously, it is necessary, as a general rule, to form in the reactor a bath of molten glass, usually at a temperature of between 1000 and 1600° C., preferably at a temperature above 1200° C., and advantageously above 1400° C., typically between 1400 and 1500° C. This bath is generally obtained using a preliminary step consisting in introducing only vitrified matrix precursors into the reactor and in subjecting them to a heat treatment step at 1400–1500° C., this generally being carried out under the conditions of step (C) defined above. Advantageously, this glass bath is obtained by induction, preferably under the conditions of forming a “cooled self-crucible”, as defined above. After the prior formation of this glass bath, constituents (i), (ii) and (iii) are added continuously, at constant flow rates, and the molten glass mixture obtained is made to flow continuously. Apart from during the start-up phase of the process, the glass composition obtained from the reactor is generally homogeneous and constant. In the context of a process carried out continuously, it is usually preferable for the ratio of the mass flow rate of metallic sodium introduced to the mass flow rate of vitrified matrix precursor introduced to be between 0.03 and 0.3, advantageously this mass ratio is less than 0.2, and preferably less than 0.14. Typically, it is between 0.05 and 0.12. The ratio of the molar flow rate of iron introduced in the form of iron oxide Fe2O3 to the molar flow rate of sodium introduced is itself preferably between 0.5/1 and 3/1, and it is advantageously greater than or equal to 0.6/1, and preferably greater than or equal to 0.9/1. It is also preferable for this ratio of the Fe/Na molar flow rates to be less than 2/1 and advantageously less than 1.5/1. The ratio of the mass flow rate of Fe2O3 introduced to the mass flow rate of vitrified matrix precursor introduced is also generally between 0.05 and 0.5. This ratio is preferably less than 0.35, and preferably less than 0.3. In the context of a process carried out continuously, the inflows of constituents (i), (ii) and (iii) and the outflow of the mixture in the molten state leaving the reactor may vary quite widely, depending on the amount of sodium that it is desired to treat. However, the process of the invention may achieve relatively high outflows of the molten glass composition, typically between 0.4 and 1.5 metric tons per hour. In this context, the inflows of the various constituents are generally from 15 to 160 kg per hour in the case of sodium, 60 to 540 kg per hour in the case of iron oxide Fe2O3 and 325 to 800 kg per hour in the case of the vitrified matrix precursor. As has already been emphasized, the process of the invention is most particularly suitable for immobilizing in glass form metallic sodium containing radioactive elements, such as the liquid sodium that has been employed in a primary or secondary cooling system of a fast-neutron nuclear reactor. In this context, the process of the invention allows the containment, within the vitrified matrix, of the radioactive elements present in the sodium. In the specific context of a process carried out in the presence of radioactive elements, it is generally preferable, so as to ensure optimum sodium containment, to carry out the process in a reactor maintained under a reduced pressure, generally a pressure reduced by a few hundred Pa (a few millibars) with respect to atmospheric pressure. This reduced pressure in the reactor falls within the context of what are called “dynamic containment” measures advantageously employed in the context of the treatment of radioactive waste, which measures are aimed at preventing any release of waste into the external environment (in the event of accidental leakage, what is observed is transfer from the external environment towards the contaminated medium and not in the other direction). Maintaining a reduced pressure certainly entails very low gaseous effluent levels, but these minor effluents can be controlled in a simple and effective manner by standard means widely used in the field of the treatment of radioactive gaseous effluents, such as filters of the “ultrahigh efficiency” category or else active carbon filters. According to one specific method of implementing the process of the invention, the metallic sodium (ii) may be introduced during step (A) together with another alkali metal in metallic form, generally metallic potassium. If necessary, the sodium (ii) is generally part of an Na/K mixture, usually in liquid form. This Na/K mixture may especially be a mixture of the type of those used in nuclear plant cooling systems and it may therefore contain radioactive elements. In this specific context, the process of the invention, advantageously carried out continuously, makes it possible to immobilize all the alkali metals present, and also the radioactive elements that they may possibly contain, in the form of a glass. When an additional alkali metal is present, the various conditions indicated above regarding the various mass and molar ratios that are preferably to be respected in order to carry out the process under optimum vitrification conditions are, however, to be adapted by transposing the data for sodium alone to the combination of alkali metals present. Thus, when operating with a mixture of several alkali metals, the total content of alkali metals incorporated into the final matrix will advantageously be less than 30% by mass and the total mass of alkali metals introduced, including the sodium (ii), will preferably be between 3 and 25% by mass with respect to the mass of precursor (i). Moreover, the amount of iron oxide (iii) employed is generally such that the molar ratio of the amount of iron introduced to the amount of alkali metals in metallic form, including the metallic sodium (ii), is between 0.5/1 and 3/1. Whatever its specific method of implementation, it should be emphasized that the process of the invention may be carried out within a small reactor. Thus a vitrification reactor for implementing the process of the invention may be installed directly on a site during dismantling, where it may be used to treat the sodium, generally directly coming from the cooling systems, without firstly having to transport this sodium, thereby further minimizing the risk of contaminating the external environment. In this context, the process may be employed under extremely rigorous containment conditions, compatible with the current legislation regarding the treatment of nuclear waste. Moreover, because of its simplicity of operation and the relatively inexpensive chemical compounds that it employs, the process of the invention generally requires very low operating costs. Furthermore, provided that the nature and the proportions of the various constituents (i), (ii) and (iii) are suitably adapted, the process of the invention allows vitrified matrices to be obtained that are homogeneous in composition, stable over time, mechanically strong and barely subject, if at all, to sodium leaching and devitrification, making them the matrices of choice for the containment of sodium containing radioactive elements in the form of glass packages. The specific use of the process described above, and in particular of such a process carried out continuously, for the containment, in a vitrified matrix, of radioactive elements present within metallic sodium or within a mixture of alkali metals including metallic sodium, constitutes a particular aspect of the present invention.
	description	FIG. 1 shows an improved 30B cylinder 10 constructed in accordance with the present invention. The cylinder 10 is shown inside the bottom half of a protective shipping package or xe2x80x9coverpackxe2x80x9d 12. The overpack 12 is shown supported in a cradle 8 and with its top half removed and its safety straps open. As is well understood in the art, during shipment to cylinder 10 is filled with up to 5,020 pounds of substantially pure uranium hexafluoride and fully enclosed in the overpack, as shown in FIG. 1A. For the most part the improved 30B cylinder 10 of the present invention is entirely conventional and will be described in detail only in so far as it differs from the prior art conventional cylinder. The conventional 30B cylinder 10 is manufactured according to ANSI N14.1 and ASME Boiler and Pressure Vessel Code, Section VIII, Division 1. Accordingly the conventional 30B cylinder is 81xc2xd inches plus or minus xc2xd inch and has a diameter of 30 inches plus or minus xc2xc inch. The conventional 30B cylinder has a minimum volume of 26 cubic feet. It is preferred that the cylinder be manufactured according to ANSI N14.1-2000 and therefore include the advantages described in U.S. Pat. No. 5,777,343 which stem from the elimination of a weld backing bar. However, the advantages of the present invention may also be obtained with cylinders manufactured to earlier versions of ANSI N14.1 which required weld backing bars. The improved 30B cylinder 10 includes a valve which is protected by a valve protection cover assembly 14 (FIGS. 1 and 2). This cover assembly, not found in conventional 30B cylinders, provides a second barrier to the egress of uranium hexafluoride or, more critically, the ingress of water. The valve protection cover assembly 14 fits within the chime 15 which extends from the domed end or head of the cylinder 10. More particularly, the distal end of the valve protection cover assembly 14 is recessed at least xc2xd inch and preferably 0.75 inches or more from the plane defined by the free edge of the chime. This space allows for deformation of the overpack during the drop test without any contact with the valve protection cover assembly 14. Therefore the cylinder 10 fitted with the cover assembly 14 may be used with standard overpacks such as the overpack 12 shown in FIGS. 1 and 1A. It should be noted that the axial length of the chime 15 is not fixed by ANSI N14.1, but the overall length, the diameter, and the minimum capacity for the cylinder are fixed. The diameter and length are critical dimensions to ensure that a tank fits in a conventional overpack. Until applicants"" invention it had not been recognized that lengthening one chime 15 and shortening the other (unnumbered) to allow a xc2xd to xc2xe or greater inch clearance as discussed above would allow a valve protection cover assembly to survive a 30 foot drop test undamaged, indeed untouched, by the deformation of the overpack, this despite the improved safety and likely resulting reduction in transportation index. The valve protection cover assembly 14 (FIG. 2) includes a cap 16 that is held in place by six bolts 18. Two of the bolts 18 are safety wired, and the wire is sealed to guarantee that the cap 16 has not been tampered with once it is bolted in place. Additional bolts, up to all six, could be safety wired if desired. The valve protection cover assembly 14, as shown in greater detail in FIG. 4, includes a cap 16 and a base 20. The base 20 is an annular disk that surrounds the valve 30. The base 20 is a disk that is welded to the wall 22 of the cylinder 10. Its diameter and thickness are selected so as not to interfere with the standard industry plumbing used to connect with the valve 30 to fill or empty the cylinder 10 of uranium hexafluoride. The base 20 is welded to the wall 22 continuously around its outer and inner-perimeters, and these welds are thoroughly inspected to guarantee their integrity. These welds therefore provide a reliable barrier to prevent any matter from passing under the base 20 and so passing from the outside of the cylinder 10 into the volume where the cap assembly surrounds the valve 30 or vice versa. The base 20 also includes six evenly spaced threaded bores (not shown) with which the bolts 18 cooperate to hold the cap 16 in place. An upper surface 24 of the base 20 includes two regions, an inner region 28 and an outer region 30. The inner region 28 is annular and stands proud of the outer region by about {fraction (1/32)} inches. The inner region 28 is machined flat and provides a working surface against which the cap 16 seals. The necessary surface flatness may be achieved by machining the base 20 either before or after welding the base 20 to the wall 22. The cap 16 is a fabricated steel component which includes a dome 40 and a flange 42. While cap 16 could be machined from a single piece of steel, it is preferred for economy and ease of manufacture to fabricate it from two pieces which are welded together as shown. This weld is thoroughly inspected to guarantee its integrity. The flange 42 mates with the base 20. To this end the flange 42 includes a machined annular surface 44 which seats against the corresponding flat inner surface 28 of the base 20. A pair of O-rings 46 and 48 fit in recesses 50 and 52, respectively, which are formed in the annular surface 44 of the flange 42. The recesses 50 and 52 are circular in plan view, but any endless shape could be used if desired. The recesses 50 and 52 may be formed with a slight undercut as shown in order to retain the O-rings 46 and 48 in place. When the annular surface 44 and the annular surface 28 are seated against each other, the O-rings 46 and 48 are compressed to form an effective seal. This seal is sufficiently complete to achieve a leak rate of less than 10xe2x88x923 ref.-cm3/sec, when tested according, for example, to the soap bubble test described in A.5.7 of ANSI N14.5-1997, Leakage Tests on Packaging for Shipment. Under this test a xe2x80x9creference cubic centimeter cubed per secondxe2x80x9d is defined as a volume of one cubic centimeter of dry air per second at one atmosphere absolute pressure and 25xc2x0 C. A seal which has the above leak rate or less is considered essentially impermeable for purposes of this application. While conventional O-rings 46 and 48 are preferred for ease of manufacture, other resilient sealing elements including cast-in-place rubbers or resilient polymers such as urethane are also possible. Such alternative materials and manufacturing techniques need only provide a sufficiently leak resistant seal to be satisfactory, and they are included within the meaning of the term xe2x80x9cresilient seal elementsxe2x80x9d used in this application. The flange 42 includes an annular outer region 58, recessed from the plane of annular surface 44. The outer region 58 is aligned with the outer region 30 of the base 20 . The two outer regions 30 and 58 define a gap 60 between them when the cap 16 is in place on the base 20. The flange 42 has six holes (not shown) through the outer region 58 for the bolts 18. These holes aligned with corresponding threaded passages in the base 20. When the cap 16 is put in place and the bolts 18 tightened to a predetermined torque, the outer region 58 of the flange 42 is stressed, assuring a predetermined, constant load on the O-rings 46 and 48 and the mating annular surfaces 24 and 44. While forming the gap 60 is preferred because it allows the flange 42 to flex slightly, any design that allows a sufficiently tight seal between the base 20 and the cap 16 is acceptable. The valve protection cover assembly 14 includes a means for testing the integrity of the seal between the cap 16 and the base 20. This test facility includes a test port 60, which leads through internal passages 62, 64, and 66 to test channel 68. The test channel 68 is a semicircular recess (in vertical cross-section) in the annular surface 44 of the flange 42. The recess 68 extends in a complete circle spaced between the recesses 50 and 52. The flange 42 includes a bore 70 (FIGS. 1 and 4) diametrically opposite the test port 60. This bore cooperates with a pin 72 which projects up from the outer region 28 of the base 20. When the cylinder 10 is in its normal, horizontal position, the pin 72 is at the 12 o""clock position and helps the worker accurately position the cap and place the bolts 18 in their holes. Once the cap 16 is in place and the bolts 18 tightened appropriately, the integrity of the seal around about may be tested. This is done by connecting the test port to a calibrated source of fluid under pressure or vacuum. The fluid reaches the test channel 68, and if the joint is secure, the fluid can go no farther. If a leak occurs, then the test equipment shows a drop in pressure or vacuum, and the O-ring seals can be inspected and replaced or other repairs made as necessary. Once the testing is complete, a plug 70 is used to seal off the test port. There are a variety of test procedures available, and these are set out in ANSI N14.5-1977. These tests assure leakage rate equal to or less than 1xc3x9710xe2x88x923 ref-cm3/sec. Although the testing facility is shown as a port, passages, and channel machined in the flange 42 of the cap 16, it is also possible to machine these elements into the base 20. If this is done, the test channel is formed in the surface 28 of the base 20 so that it is located between the places where the O-rings contact the base 20 and is connected to a test port by suitable passages. Similarly, the O-rings 46 and 48 could be mounted in grooves formed in the base. However, the construction shown in the Figures is preferred because it is easier to maintain and because the O-rings 46 and 48 and the test channel 68 are less likely to be damaged when connecting conduits the valve 30. While the bolts 18 are used to draw the cap 16 tight against the base 20, other fastenings are possible. For example a threaded connection between the base could be used with the necessary O-ring seals and test port channel formed in a screw-on cap. Alternatively, the base 20 could have external threats on its outer peripheral surface and a nut like that used in a plumber""s union could be used to pull the cap down against the base. Thus it is clear that the present invention provides a vessel 10 for the shipment of uranium hexafluoride includes a cylindrical wall closed by pair of approximately semi-ellipsoidal heads 22 welded to form a sealed container. A service valve 30 is located in one end. The valve 30 is covered by a removable, watertight valve protection cover assembly 14. The vessel also includes a test port 60 by means of which the integrity of the valve protection cover assembly may be tested after the cylinder 10 has been filled with uranium hexafluoride and the valve protection assembly 14 has been installed. The valve protection assembly 14 is shaped so that it fits within the envelope of the standard 30B cylinders, and so fits within the overpacks already approved by the NRC and owned by shippers of uranium hexafluoride. The vessel 10 made according to the present invention has a double barrier to prevent ingress of water or egress of uranium hexafluoride. The valve 30, a first barrier, is enclosed by a cover assembly 14 which forms the second barrier. The double barrier is expected to permit the transportation index of 0. In effect, then, adding the second barrier will allow the improved 30B cylinders to be shipped in bulk with safety acceptable to the NRC, resulting in substantial savings to the industry.
044302568	description	DETAILED DESCRIPTION The safe disposal of nuclear waste has exercised the concern of the entire scientific community. As shown in FIG. 1 at present the system relies on two separate major components in the system to prevent radionuclides from reaching the biosphere 6, from the waste package 2, through the geological host rock 4. FIG. 1 shows the waste package 2 in repository designed to be maximally insoluble, so as not to release radionuclides under repository conditions and the geological host rock 4 selected to make travel of radionuclides to the surface maximally slow and difficult. The present invention is concerned with barrier 2. If barrier 2 is looked at in greater detail, this can be conceived as a series of sequential component barriers (by analogy to the series of nested Russian dolls) each barrier offering resistance to the release of ions from the total package. This waste package 2 then consists of four components, as shown in FIG. 2, namely the waste form (radio phase+encapsulant) 8, the canister 10, overpack 12, and liner 14. Previously, virtually all of the research and invention on the waste package has focused on 8 the waste form, with some attention on 10 the canisters. The design objective has been to make these materials maximally insoluble in the repository environment. The invention is directed to the barrier 12, the overpack (also sometimes called the backfill). Previous concern with this component has been shown in the Swedish KBS plan, where a mixture of bentonite, quartz, and ferrous phosphate was proposed. The present inventor and his colleagues in the past also have presented various papers in which they have designed new materials for this barrier. The goal of all the art to date, regarding the function of this overpack barrier 12 is shown in connection with FIG. 3. Up to the present, this barrier has been conceived as a means either excluding water or of absorbing (and reacting with, in some degree) the dangerous radionuclides, chiefly Cs, Sr, and the actinides (generic symbol, An). Illustrative of the actinides with which the present invention is concerned are Th, U, Pu, Am, Np and Pa. Other important radionuclides include I and Tc. The fundamental innovation in the present invention is to reverse the chemical concentrations so that there will be a tendency for Cs, Sr, I, Tc, and An ions or in place of An ions optionally Ln (lanthanide) ions to move into the waste form rather than the other way. This is achieved by changing the gradient of the chemical potential of the elements of concern, so that there will be a minimum chemical driving force for any of the dangerous species to migrate outward. This reverse chemical potential gradient is very easily attained. It simply requires that the overpack be saturated with minerals or chemicals containing Sr, Cs, Ln, etc., which are more soluble than the waste form in the repository fluids. This is shown in FIG. 4. The only meaningful threat comes from transport and reaction in fluids since solid state diffusion rates are much too slow to be of concern. The elements of concern only enter these solutions as ions. Diffusion processes in the liquid will move ions from high concentration areas to low concentration areas. Thus, one creates a positive chemical potential gradient for these ionic species towards the waste form, if and when there is a breach of the containment and material can flow (ever so slowly) in or out of the canister. In lieu of the aforesaid "overpack" container technique is the backfill technique. That is to say, the earth surrounding a "normal" canister (a container not characterized by an overpack material) comprises non-radioactive ions of higher activity than the same radioactive ions stored in the "normal" canister. By this backfill technique, pernicious radioactive ions eventually escape from the container but are "prevented from diffusing further" by the same non-radioactive ions comprising the backfill. In other words, the overpack or backfill in this invention is designed not only to absorb the radionuclides of the dangerous elements Cs, Sr, An, I, Tc originating in the waste form, but it is also designed to block the out migration of any such ions by providing a supply of non-radioactive atoms of the same elements outside the canister, mixed into the overpack. This layer of non-radioactive (hence not dangerous) atoms of the same elements serves as a highly impenetrable chemical or thermodynamic barrier. Most of the long-term threat is from the .alpha.-emitting actinides. It is easy enough to obtain non-radioactive Cs and Sr chemicals and minerals. However, except for uranium and thorium, the actinides are all laboratory rarities. Hence there is used instead as excellent imitators or substitutes the larger, lighter members of the lanthanide group. Because of the identity of the ionic radii the corresponding ions from the two series are very similar in solid state reactions. Thus, there can be used principally, La, Ce, Pr, Nd, Sm, Eu, Gd, or mixtures thereof with yttria earth in place of the actinides. In place of Tc, there can be used the much more common Mn. This invention relates to any radionuclide, and the blocking of its migration by incorporating a stable nuclide of the same element in the overpack or a replacement element such as a lanthanide or Mn. A major advantage of the present invention over the conventional efforts to make increasingly insoluble waste forms, is the fantastic complexity and expense of working with highly radioactive materials in a remote "canyon" facility. The engineering of the overpack or backfill is a matter of extreme simplicity since none of the matter is radioactive, it consists of selecting the desirable mineral (or ceramic) phase and mixing it into the overpack as loose powder to be tamped, or as formed shapes. One of the configurations in which the new material will be used is shown in FIG. 5. Here the "ion-doped" layer is incorporated as a layer of briquettes 16. Another configuration is shown in FIG. 6 where the ion-doped material is simply tamped in sequence as the overpack material is put in place. A third arrangement is simply to have the entire overpack layer contain some of the ion-doped material. While the Cs, Sr, Ln, An, I, Mn and other ions can be introduced as virtually any salt, cost-effectiveness dictates that these be added as relatively insoluble materials, only slightly more soluble in the probable repository environment than the waste form itself. Since the waste forms are designed to be maximally insoluble, one can use relatively small amounts of quite insoluble phases. Thus, for Sr, its common ore celestite (SrSO.sub.4) is adequate in some repository environments; however, in most, a ceramic material such as strontium feldspar (SrAl.sub.2 Si.sub.2 O.sub.8) or its partial solid solutions with ordinary calcium feldspar of the general formula Ca.sub.1-x Sr.sub.x Al.sub.2 Si.sub.2 O.sub.8 where x is a number less than 1 and greater than 0 will suffice. Similar series of solid solutions with the structural formula Ca.sub.1-x Sr.sub.x SiO.sub.3 --which can be made readily by reacting CaCO.sub.3, SrSO.sub.4, and sand, are also suitable. For the introduction of Cs, various Cs-containing mineral phases (natural and synthetic) are available. Among them, natural and synthetic pollucite and CsAlSi.sub.5 O.sub.12, partial solid solutions of the (Ba.sub.1-x Cs.sub.2x) variety in the celsian, magnetoplumbite, or hollandite structures. For the actinides, a mixture of various uranium and thorium minerals (such as uraninite, thorite, etc.) and rare-earth ores (such as bastnaesite), should be adequate. However, synthetic rare-earth silicates and aluminates made by reacting the "natural" mix of the larger rare-earth ions with silica or alumina, can be tailored to a solubility just slightly greater than the waste form. In summary, there is placed a special layer of overpack or backfill material around a nuclear waste canister, the special layer comprising a natural mineral and/or a ceramic material or other source of the non-radioactive analogue of the radionuclide(s) of concern. Preferably, the composition of this overpack or backfill is completely and simply adjusted by selecting and combining appropriate mineral or ceramic phases which are only slightly more soluble than the waste form in repository fluids, e.g., water. Pollucite, CsAlSi.sub.5 O.sub.12, and the appropriate solid solution of Cs in magnetoplumbite, celsian, or hollandite, and virtually all relatively insoluble Cs compounds with Al.sub.2 O.sub.3, SiO.sub.2, P.sub.2 O.sub.5, and TiO.sub.2 in any combination are useful additives for this tailored overpack. Celestite (SrSO.sub.4), strontium feldspar, or any other strontium compounds with Al.sub.2 O.sub.3, SiO.sub.2, P.sub.2 O.sub.5 singly or in any combination make excellent strontium overpack materials. All the major natural ores of uranium and thorium can serve as sources of actinides in the overpack. In addition, the rare earth ores or oxides themselves, or the combinations of them with Al.sub.2 O.sub.3, SiO.sub.2 and/or P.sub.2 O.sub.5 can provide lanthanide ions to mimic the actinides in the new tailored overpack and used in place of them. The amounts of "ionically-charged" overpack that will be used around each canister is an engineering parameter readily and easily chosen by the systems designer, just as is the thickness of the canister and dilution of the waste form. According to the invention, there is provided a process for constructing a "chemical container" or "reverse thermodymanic barrier" by surrounding a canister of nuclear waste placed in a repository with an appropriate (e.g., natural or synthetic) mineral overpack containing a substantial amount of Cs, Sr, An, and other fission product radionuclide ions. This overpack may be emplaced in one of several ways: (1) As a tamped-in mixture of clays, zeolites, etc., with the ion-dopant materials; PA1 (2) As ceramic briquettes of the ion-dopant materials surrounding an inner layer of overpack and optionally more overpack outside the briquettes; PA1 (3) In various concentrations dispersed throughout the overpack (or backfill) material. Such a reverse thermodymanic chemical gradient of species such as Cs, Sr, An, I, Te, Mn (for Tc), or Ln (for An) achieved by placing between 0.1 and 100.times. (where .times. stands for times) the total contained amount of the ion contained in the waste form into appropriately insoluble overpack materials will provide a more cost-effective total waste package than engineering a more highly insoluble waste form. The upper limit of 100.times. is not critical, and much higher amounts can be used in the overpack but are not normally justified economically. Criticality of the lower limit is that there be sufficient ions present in the overpack to insure a tendency of the ions of the elements involved to go into the waste form or other container. Thus, in a Batelle process using a glass matrix the release rate of the radioactive material is about 10.sup.-5 gm/cm.sup.2 per day and the ion concentration in the overpack need only be sufficient to overcome this gradient and prevent waste from going through the canister. In other words, the concentration of ions in the overpack should be sufficient to exceed the leachability rate into the environment. Usually the concentration of ions in the overpack will exceed 1.times. the concentration of radionuclide ions in the waste. The product can comprise, consist essentially of, or consist of the stated materials and the process comprise, consist essentially of, or consist of the recited steps with the stated materials.
054208971	abstract	A fast reactor comprises a core composed of nuclear fuel, a core barrel surrounding an outer periphery of the core, an annular reflector surrounding an outer periphery of the core barrel, a partition wall structure surrounding an outer periphery of the annular reflector and supporting the core barrel by a supporting structure arranged radially of the fast reactor, the partition wall structure constituting an inner wall of a coolant passage for a primary coolant, a neutron shield surrounding an outer periphery of the partition wall structure and disposed in the coolant passage, a reactor vessel surrounding an outer periphery of the neutron shield and having an inner wall constituting an outer wall of the coolant passage, and a guard vessel surrounding an outer periphery of the reactor vessel.
	claims	1. A system for delivering very high electron energy beam to a targeted tissue in a patient, comprising:an accelerator capable of generating a very high electron energy beam, wherein the very high electron energy beam comprises an electron energy beam between 50 and 250 MeV;a beam steering device capable of receiving the very high electron energy beam from the accelerator and steering the beam to the targeted tissue from multiple directions; anda controller capable of controlling length of time that the very high electron energy beam irradiates the targeted tissue in delivering an entire treatment dose, the length of time of delivering the entire dose being less than 10 seconds, and to control the directions in which the beam steering device steers the beam to the targeted tissue when delivering the dose. 2. A system according to claim 1 wherein the controller is configured to receive information from an imaging or targeting device and use the information from the imaging device to control the directions in which the beam steering device steers the beam to the targeted tissue, wherein the entire is delivered to the targeted tissue by the beam steering device based on position verification data received from the imaging or targeting device. 3. A system according to claim 1 wherein the accelerator is capable of generating a beam of between 75 and 100 MeV. 4. A system according to claim 1 wherein the length of time of delivering the entire dose is less than one second. 5. A system according to claim 1 wherein the beam steering device is selected from the group consisting of electro-magnetic devices and radiofrequency deflector devices. 6. A system according to claim 1 wherein the beam steering device includes a gantry, the gantry including multiple beam ports through which the beam steering device can steer the very high electron beam received from the accelerator. 7. A system according to claim 1 wherein the beam steering device is capable of providing thin pencil beam raster scanning. 8. A system according to claim 1 wherein the beam steering device includes a continuous annular gantry. 9. A system according to claim 1 wherein the beam steering device is capable of providing volume filling scanning. 10. A system according to claim 1 wherein the beam steering device is configured to steer the very high electron energy beam from multiple directions without requiring movement of any mechanical parts. 11. A system for delivering very high electron energy beam to a targeted tissue in a patient, comprising:an accelerator capable of generating a very high electron energy beam;a beam steering device capable of receiving the very high electron energy beam from the accelerator and steering the beam to the targeted tissue from multiple directions;a controller capable of controlling length of time that the very high electron energy beam irradiates the targeted tissue in delivering an entire dose of radiation to the targeted tissue, the length of time of delivering the entire dose being less than 10 seconds, and to control the directions in which the beam steering device steers the beam to the targeted tissue; andan imaging or targeting device capable of locating the targeted tissue and providing location information to the controller configured to control the directions in which the beam steering device steers the beam to the targeted tissue. 12. A system according to claim 11 wherein the imaging or targeting device is capable of providing information to the controller of a location of the targeted tissue so as to trigger when the system delivers the beam to the targeted tissue at the location. 13. A system according to claim 11 wherein, using information from the imaging or targeting device, the system is capable of automatically delivering the beam to the targeted tissue from multiple predetermined directions at multiple predetermined points in time completing delivery of the entire dose from all directions within less than 10 seconds. 14. A system according to claim 11 wherein the accelerator is capable of generating a beam of between 75 and 100 MeV. 15. A system according to claim 11 wherein the length of time of delivery of the entire dose by irradiation of the targeted tissue with the beam is less than one second. 16. A system according to claim 11 wherein the beam steering device is selected from the group consisting of electro-magnetic devices and radiofrequency deflector devices. 17. A system according to claim 11 wherein the beam steering device is configured to steer the very high electron energy beam from multiple directions without requiring movement of any mechanical parts. 18. A method for delivering a beam of very high electron energy to a targeted tissue in a patient, comprising:providing a system for delivering very high electron energy beam to a targeted tissue in a patient, the system comprising:an accelerator capable of generating a very high electron energy beam;a beam steering device capable of receiving the very high electron beam from the accelerator and steering the very high electron beam to the targeted tissue from multiple directions;a controller capable of controlling a length of time that the very high electron beam irradiates the targeted tissue to deliver an entire treatment dose, the length of time being less than 10 seconds, and to control the directions in which the beam steering device steers the beam to the targeted tissue when delivering the dose; andactuating the system to cause it to deliver the beam to the targeted tissue. 19. A method according to claim 18 wherein actuating the system causes a beam of between 75 and 100 MeV to be generated by the accelerator. 20. A method according to claim 18 wherein the length of time of delivery of the entire dose by irradiation of the targeted tissue with the beam is less than one second. 21. A method according to claim 18 wherein providing the system includes providing a beam steering device that is selected from the group consisting of electro-magnetic devices and radiofrequency deflector devices. 22. A method according to claim 18 wherein providing the system includes providing a beam steering device that steers the beam to the targeted tissue from multiple directions without requiring movement of any mechanical parts. 23. A method according to claim 18 further comprising providing a controller that is configured to receive information from an imaging or targeting device as to a position of the targeted tissue and use the information from the imaging or targeting device to control the directions in which the beam steering device steers the beam to the targeted tissue at the determined position and complete delivery of the entire dose before substantial movement of the targeted tissue from the determined position. 24. A method according to claim 18 further comprising providing an imaging device capable of generating images of the targeted tissue and providing information from the imaging device to the controller to control the directions in which the beam steering device steers the beam to the targeted tissue. 25. A method according to claim 24 wherein providing the imaging device includes providing an imaging device that is capable of providing information to the controller to trigger when the system delivers the beam to the targeted tissue. 26. A method according to claim 24 wherein providing the imaging device includes providing an imaging device wherein, using information from the imaging device, the system is capable of automatically delivering the beam to the targeted tissue from multiple predetermined directions concurrently or in rapid succession at multiple predetermined points in time completing delivery of the entire dose from all directions within less than 10 seconds. 27. A system for delivering a transverse-modulated electron beam to a targeted tissue in a patient, comprising:a photoelectron gun configured to generate a transverse-modulated electron beam from an optical image produced by a light source and projected or scanned on a photocathode;an accelerator capable of increasing the energy level of the transverse-modulated electron beam to a predetermined level;a beam steering device capable of receiving the transverse-modulated electron beam from the accelerator and steering the transverse-modulated electron beam to the targeted tissue from multiple directions; anda controller capable of controlling length of time that the transverse-modulated electron beam irradiates the targeted tissue in delivering an entire treatment dose, the length of time being less than 10 seconds, and to control the directions in which the beam steering device steers the transverse-modulated electron beam to the targeted tissue when delivering the dose. 28. A system for delivering a transverse-modulated electron beam to a targeted tissue in a patient according to claim 27 wherein the light source is a laser. 29. A system for delivering a transverse-modulated electron beam to a targeted tissue in a patient according to claim 27, wherein the optical image comprises an intensity pattern corresponding to a dose distribution planned for delivery to the targeted tissue based on a given shape of the targeted tissue and further comprising:an optics system capable of maintaining the intensity pattern with high fidelity during steering of the transverse-modulated beam to the targeted tissue with the beam steering device. 30. A system for delivering a transverse-modulated electron beam to a targeted tissue in a patient according to claim 27, wherein the controller is configured to:produce a first 2-dimensional intensity pattern with the transverse-modulated electron beam with the photo-electron gun and accelerator and to steer the beam to irradiate the targeted tissue with the first 2-dimensional pattern from a first direction, andproduce one or more additional 2-dimensional intensity patterns with the transverse-modulated electron beam with the photo-electron gun and accelerator and to steer the beam to irradiate the targeted tissue with the one or more additional 2-dimensional patterns from one or more additional directions, respectively, such that summing the first 2-dimensional intensity pattern and the one or more additional 2-dimensional patterns across the first direction and the one or more additional directions provides the planned dose distribution, wherein the planned dose distribution is a 3-dimensional dose distribution and the given shape is 3-dimensional.
	claims	1. A method for assessing a quality of a digital slide image, the method comprising using at least one hardware processor to:divide a digital slide image into a plurality of image regions;for each of at least a subset of two or more of the plurality of image regions,determine a spatial frequency of the image region, anddetermine a quality of the image region based on the determined spatial frequency; andgenerate a visual depiction of the digital slide image that, for each of the at least a subset of two or more of the plurality of image regions, indicates the determined quality of the image region. 2. The method of claim 1, further comprising using the at least one hardware processor to, for each of the plurality of image regions, determine whether the image region has qualified content, wherein the at least a subset of two or more of the plurality of image regions comprises only those image regions of the plurality of image regions that have been determined to have qualified content. 3. The method of claim 1, wherein determining a spatial frequency of the image region comprises performing one or more frequency transforms on the image region to determine a score for the image region, and wherein the quality of the image region comprises the score for the image region. 4. The method of claim 3, wherein performing one or more frequency transforms on the image region to determine a score for the image region comprises:performing a plurality of partial frequency transforms on the image region to generate a plurality of partial frequency transform values; andgenerating a score for the image region based on one or more of the plurality of partial frequency transform values. 5. The method of claim 4, wherein generating a score for the image region based on one or more of the plurality of partial frequency transform values comprises generating the score for the image region by combining those of the plurality of partial frequency transform values that exceed a threshold value. 6. The method of claim 1, further comprising using the at least one hardware processor to generate a score for the digital slide image based on the determined quality of one or more of the at least a subset of two or more of the plurality of image regions. 7. The method of claim 6, further comprising using the at least one hardware processor to:detect artifacts in the digital slide image based on the determined quality of one or more of the at least a subset of two or more of the plurality of image regions; anddetermine a degradation score based on one or more of a magnitude and proportion of detected artifacts;wherein the score for the digital slide image is further based on the determined degradation score, andwherein the visual depiction of the digital slide image comprises a visual identification of each detected artifact. 8. The method of claim 6, further comprising using the at least one hardware processor to, if the score for the digital slide image is below a threshold value, provide an indication that a slide corresponding to the digital slide image needs to be inspected or rescanned. 9. The method of claim 1, wherein the visual depiction of the digital slide image comprises each of the plurality of image regions with each of the at least a subset of two or more of the plurality of image regions depicted in one of a plurality of colors that corresponds to the determined quality of that image region, wherein each of the plurality of colors represents a different level of quality. 10. A system for assessing a quality of a digital slide image, the system comprising:at least one hardware processor; andone or more modules that are configured to, when executed by the at least one hardware processor,divide a digital slide image into a plurality of image regions,for each of at least a subset of two or more of the plurality of image regions, determine a spatial frequency of the image region, and determine a quality of the image region based on the determined spatial frequency, andgenerate a visual depiction of the digital slide image that, for each of the at least a subset of two or more of the plurality of image regions, indicates the determined quality of the image region. 11. The system of claim 10, wherein the one or more modules are further configured to, for each of the plurality of image regions, determine whether the image region has qualified content, wherein the at least a subset of two or more of the plurality of image regions comprises only those image regions of the plurality of image regions that have been determined to have qualified content. 12. The system of claim 10, wherein determining a spatial frequency of the image region comprises performing one or more frequency transforms on the image region to determine a score for the image region, and wherein the quality of the image region comprises the score for the image region. 13. The system of claim 12, wherein performing one or more frequency transforms on the image region to determine a score for the image region comprises:performing a plurality of partial frequency transforms on the image region to generate a plurality of partial frequency transform values; andgenerating a score for the image region based on one or more of the plurality of partial frequency transform values. 14. The system of claim 13, wherein generating a score for the image region based on one or more of the plurality of partial frequency transform values comprises generating the score for the image region by combining those of the plurality of partial frequency transform values that exceed a threshold value. 15. The system of claim 10, wherein the one or more modules are further configured to generate a score for the digital slide image based on the determined quality of one or more of the at least a subset of two or more of the plurality of image regions. 16. The system of claim 15, wherein the one or more modules are further configured to:detect artifacts in the digital slide image based on the determined quality of one or more of the at least a subset of two or more of the plurality of image regions; anddetermine a degradation score based on one or more of a magnitude and proportion of detected artifacts;wherein the score for the digital slide image is further based on the determined degradation score, andwherein the visual depiction of the digital slide image comprises a visual identification of each detected artifact. 17. The system of claim 15, wherein the one or more modules are further configured to, if the score for the digital slide image is below a threshold value, provide an indication that a slide corresponding to the digital slide image needs to be inspected or rescanned. 18. The system of claim 10, wherein the visual depiction of the digital slide image comprises each of the plurality of image regions with each of the at least a subset of two or more of the plurality of image regions depicted in one of a plurality of colors that corresponds to the determined quality of that image region, wherein each of the plurality of colors represents a different level of quality. 19. A non-transitory computer-readable medium having one or more sequences of instructions stored thereon, wherein the one or more sequences of instructions are configured to, when executed by at least one hardware processor, cause the at least one hardware processor to:divide a digital slide image into a plurality of image regions;identify at least a subset of two or more of the plurality of image regions having qualified content;for each of the at least a subset of two or more of the plurality of image regions,perform a plurality of partial frequency transforms on the image region to generate a plurality of partial frequency transform values, andcombine those of the plurality of partial frequency transform values that exceed a threshold value to generate a score for the image region;generate a score for the digital slide image based on the generated score of one or more of the at least a subset of two or more of the plurality of image regions; andgenerate a visual depiction of the digital slide image that comprises each of the plurality of image regions with each of the at least a subset of two or more of the plurality of image regions depicted in one of a plurality of colors that corresponds to the determined quality of that image region, wherein each of the plurality of colors represents a different level of quality. 20. The non-transitory computer-readable medium of claim 19, wherein the one or more sequences of instructions are further configured to, when executed by at least one hardware processor, cause the at least one hardware processor to:detect artifacts in the digital slide image based on the determined quality of one or more of the at least a subset of two or more of the plurality of image regions; anddetermine a degradation score based on one or more of a magnitude and proportion of detected artifacts;wherein the score for the digital slide image is further based on the determined degradation score, andwherein the visual depiction of the digital slide image further comprises a visual identification of each detected artifact.
051046091	claims	1. An assembly method for a nuclear fuel assembly, comprising the steps of: preparing at least one first grid member with springs on wall sections thereof, and preparing at least one second grid member with dimples on wall sections thereof; arranging the first grid member and the second grid member face to face such that a plurality of grid cells of the first grid member and a plurality of grid cells of the second grid member communicate with each other and that said dimples of the second grid member and said springs of said first grid member are disposed on parallel planes which are opposed to each other; providing a sleeve on one of said first and second grid members for allowing lateral shifting of said first grid member with respect to said second grid member, retaining the first and second grid members by a grid retainer, inserting a guide thimble into a thimble cell of the first and second grid members while retaining a clearance between the first and second grid members, fixing the guide thimble to the second grid member, inserting fuel rods respectively into the grid cells of the first grip member and the grid cells of the second grid member such that the springs of said grid member and the dimples of the second grid member are shifted relative to each other in such a direction that the springs and the dimples move away from each other; subsequently moving at least one of the first grid member and the second grid member such that the grid cells in the first grid member and the grid cells in the second grid member are in alignment with each other; and connecting the first grid member and the second grid member to each other. 2. An assembly method according to claim 1, including preparing a pair of said at least one second grid members, arranging said first grid member and said pair of second grid members such that the three grid members are superimposed on each other with said first grid member interposed between said pair of second grid members, and connecting said superimposed three grid members to each other by an outer frame. 3. An assembly method according to claim 2, wherein each of said three grid members includes a rectangular outer strap and a plurality of inner straps, and which comprises mounting said outer strap to four sides corresponding to ends of said innerstraps which are assembled together into a grid form to define the grid cells, each of said second grid members having a plurality of pairs of dimples, each of said pair of dimples being provided respectively on adjacent two of four wall sections formed by the inner and outer straps of the second grid member, said four wall sections defining one of the grid cells of each of the second grid members, said first grid member having a plurality of pairs of springs, each pair of springs being provided respectively on adjacent two of four wall sections formed by the inner and outer straps of the first grid member with springs, and said four wall sections defining one of the grid cells of the grid member with springs. 4. As assembly method according to claim 3, including, prior to inserting the fuel rods, mounting a part of the outer frame onto a Y-shaped support, mounting an assembly of said three grid members onto the Y-shaped support, and after inserting the fuel rods into the respective grid cells of the pair of second grid members and the grid cells of the first grid member, moving the first grid member such that the grid cells in the pair of second grid members and the grid cells in the first grid member are aligned with each other, and fixing the other part of the outer frame to the assembly of the three grid members and to said one part of the outer frame. 5. An assembly method according to claim 1, including arranging a pair of the firs grid members and a pair of the second grid members such that the four grid members are superimposed on each other with the pair of first grid members interposed between the pair of second grid members, and connecting the superimposed four grid members to each other by an outer frame. 6. An assembly method according to claim 5, wherein each of said four grid members comprises an outer strap and a plurality of inner straps, said outer strap being mounted to four sides corresponding to the ends of said inner straps which are assembled together into a grid form to define the grid cells, each of said pair of second grid members having a plurality of pairs of dimples each pair provided respectively on adjacent two of four wall sections of the inner and outer straps of the second grid members, said four wall sections defining one of the grid cells of the second grid members, each of said pair of first grid members having a plurality of springs each provided on one of four wall sections of the inner and outer straps of the first grid members, said four wall sections defining one of the grid cells of the first grid members, and wherein the springs of the pair of first grid members are arranged in facing relation to the dimples of said pair of second grid members. 7. An assembly method according to claim 6, including, prior to inserting the fuel rods, retaining the four grid members onto a grid retainer, inserting a guide thimble into thimble cells of the respective four grid members, while retaining slight clearances respectively between the four grid members, bulge-fixing the guide thimble to the respective pair of second grid members, mounting one part of the outer frame onto a Y-shaped support, and, after mounting an assembly of said four grid members onto the Y-shaped support, inserting the fuel rods respectively into the grid cells of the pair of second grid members and the grid cells of the pair of first rid members, moving the pair of first grid members such that the grid cells in the pair of second grid members and the grid cells in the pair of first grid members are aligned to each other, and fixing the other part of the outer frame to the assembly of the four grid members and to said one part of the outer frame.
	abstract	A control rod drive system (CRDS) for use in a nuclear reactor. In one embodiment, the system generally includes a drive rod mechanically coupled to a control rod drive mechanism (CRDM) operable to linearly raise and lower the drive rod along a vertical axis, a rod cluster control assembly (RCCA) comprising a plurality of control rods insertable into a nuclear fuel core, and a drive rod extension (DRE) releasably coupled at opposing ends to the drive rod and RCCA. The CRDM includes an electromagnet which operates to couple the CRDM to DRE. In the event of a power loss or SCRAM, the CRDM may be configured to remotely uncouple the RCCA from the DRE without releasing or dropping the drive rod which remains engaged with the CRDM and in position.
	summary
	claims	1. A corrosion resistant, thermal neutron absorbing alloy powder having the following composition in weight percent,Carbonnot more than about 0.08Manganeseup to about 3Siliconup to about 2Chromiumabout 17 to about 27Nickelabout 11 to about 20Mo + (W/1.92)up to about 5.2Boronat least about 0.25Gadoliniumat least about 0.05Oxygenabout 0.01 to about 0.1Nitrogenup to 0.2wherein BEq=% boron+(4.35×% gadolinium), 0.78≦BEq≦13.0, and the balance of the alloy powder composition is iron and usual impurities, wherein the impurities include 0.05% max. phosphorus, 0.03% max. sulfur, less than 0.005% yttrium, and less than 0.01% aluminum. 2. An alloy powder as claimed in claim 1 which contains not more than about 2.6% gadolinium. 3. An alloy powder as claimed in claim 1 which contains not more than about 2.5% boron. 4. An alloy powder as claimed in claim 1 wherein Mo+(W/1.92) is at least about 2.8%. 5. An alloy powder as claimed in claim 1 wherein Mo+(W/1.92) is not more than about 0.5%. 6. An alloy powder as claimed in claim 1 which contains not more than about 1.0% boron and BEq is not more than about 12.0%. 7. An alloy powder as claimed in claim 1 which contains not more than about 2.0% boron. 8. An alloy powder as claimed in claim 7 which contains not more than about 0.25% gadolinium. 9. An alloy powder as claimed in claim 1 which contains at least about 0.12% gadolinium. 10. An alloy powder as claimed in claim 9 which contains not more than about 2.6% gadolinium. 11. An alloy powder as claimed in claim 1 which contains at least about 300 ppm nitrogen. 12. An alloy powder as claimed in claim 1 which contains not more than about 0.05% carbon. 13. An article of manufacture that provides good processability in combination with good mechanical and corrosion resistance properties, said article being formed from consolidated austenitic alloy powder having the following composition in weight percent,Carbonnot more than about 0.08Manganeseup to about 3Siliconup to about 2Chromiumabout 17 to about 27Nickelabout 11 to about 20Mo + (W/1.92)up to about 5.2Boronat least about 0.25Gadoliniumat least about 0.05Oxygenabout 0.01 to about 0.1Nitrogenup to 0.2wherein BEq=% boron+(4.35×% gadolinium), 0.78≦BEq≦13.0, and the balance of the alloy powder composition is iron and usual impurities, wherein the impurities include 0.05% max. phosphorus, 0.03% max. sulfur, less than 0.005% yttrium, and less than 0.01% aluminum, and the article comprises a matrix and a plurality of boride and gadolinide particles dispersed within the matrix, said boride and gadolinide particles are predominantly M2B, M3B2, M5X, and M3X in form, where X is gadolinium or a combination of gadolinium and boron and M is one or more of the elements silicon, chromium, nickel, molybdenum, and iron. 14. An article as claimed in claim 13 wherein the alloy powder contains not more than about 0.05% carbon. 15. An article as claimed in claim 13 wherein the alloy powder contains at least about 0.12% gadolinium. 16. An article as claimed in claim 15 wherein the alloy powder contains not more than about 2.6% gadolinium. 17. An article as claimed in claim 13 wherein the alloy powder contains not more than about 2.0% boron. 18. An article as claimed in claim 13 wherein % Mo+(% W/1.92) in the alloy powder is at least about 2.8%. 19. An article as claimed in claim 13 wherein % Mo+(% W/1.92) in the alloy powder is not more than about 0.5%. 20. An article as claimed in claim 13 wherein the alloy powder contains not more than about 1.0% boron and BEq is not more than about 12.0%. 21. An article as claimed in claim 20 wherein the average area fraction of the borides and gadolinides is not more than about 20%. 22. An article as claimed in claim 13 wherein the average area fraction of the borides and gadolinides is not more than about 22% and the alloy powder contains not more than about 2% boron and not more than about 0.25% gadolinium. 23. An article as claimed in claim 13 wherein the alloy powder contains not more than about 1.0% boron. 24. An article as claimed in claim 23 wherein the alloy powder contains at least about 0.12% gadolinium. 25. An article as claimed in claim 24 wherein the alloy powder contains not more than about 2.6% gadolinium.
	summary
043575412	summary	The present invention relates to an apparatus for the temporary radiation-protected reception of radioactive waste of relatively short half-life. Radioactive isotopes which have a relatively short half-life of, for instance, a few hours are frequently used as indicators for diagnostic purposes in hospitals and other medical establishments. At present, 99m-Tc and 113-In are in particular use, in addition to 18-F, 123-I and 132-I. Such isotopes are generally administered by means of disposable hypodermics which contain a solution containing the corresponding isotopes. For reasons of environmental protection, the hypodermics used and other materials possibly employed, such as cotton pads, etc., which come into contact with the radioactive solution, may not simply be placed in ordinary rubbish; rather, they must be stored with radiation protection until the radioactivity fades to a safe level, over the period of a few days to a week. If radiation-protected containers are set up for the radioactive wastes of the said type which are produced each day, then the waste-filled containers must be emptied each day or replaced by an empty container, and the waste which has accumulated or the filled containers must be stored in safe manner for the required decay time, with strict monitoring of the date on which the waste was produced. This is annoying and results in a considerable loss of time. Furthermore, there is the danger that, despite the monitoring, new radioactive wastes may enter into containers in which waste has already been present for a few days or that containers will be emptied prematurely into the general rubbish. The object of the present invention is to create an apparatus for the temporary radiation-protected reception of radioactive wastes of relatively short half-life, which makes it possible, in a simple manner and without a large amount of time or space being required, to separate the periodically produced radioactive waste in accordance with the time of its origination, store it with radiation protection without further manipulation for the required period of time, and thereupon collect it for normal removal. In order to achieve this object, the invention comprises a plurality of vertical tubes, each adapted to receive a bag. The tubes are regularly spaced about the periphery of an assembled unit which is rotatable about a central vertical axis. The assembly is in a completely enclosed housing having a radiation protection jacket. A bottom plate of the housing prevents the bags from falling out of the tubes. The bottom plate and a top cover plate of the housing are each provided with an opening corresponding to the cross-sectional area of one tube. The opening of the top plate is adapted to be closed by a radiation protection lid. A reception container for the bag is adapted to be placed below the opening of the bottom plate. The openings in the top and bottom plates are offset so that when one of the tubes in the peripheral row is below the opening of the top plate, the last tube in the direction of rotation from said one tube is above the opening in the bottom plate. Maintenance of the apparatus of the invention is thus limited to turning the arrangement of tubes in the direction of rotation through the angle between two adjacent tubes, for instance at the end of each workday or at the start of the following workday and at the same time inserting a bag into the empty tube which now appears under the opening of the top plate. Upon this further rotation, the waste-filled bag in the tube, which is furthest away in the direction of rotation from the accessible tube, automatically drops through the opening of the bottom plate into the reception container for radioactive unobjectionable waste after it has spent the required period of time in radiation-protected position as a result of the individual daily rotations through the said angle.
062332996	summary	BACKGROUND OF THE INVENTION 1. The Field of the Invention The present invention relates in general to an assembly which is loaded into a fast reactor for the purposes of transmutation treatment of a long-lived radioactive material. More particularly, the present invention relates to an assembly for transmutation of a long-lived radioactive material, in which the transmutation assembly is composed solely of FP (fission product)--containing pins, in which each of the FP pins has a cladding tube containing therein a moderator and a radioactive material including long-lived fission product (LLFP) nuclides in such a manner that the radioactive material is surrounded by, or in other words covered with, the moderator material in the cladding tube. 2. Description of Prior Art Utilization of nuclear energy is inevitably followed by the generation, more or less, of long-lived fission products (LLFP). High level radioactive waste which has been disposed of by glassification generally contains long-lived radioactive FP nuclides, such as technetium-99 and iodine-129, which have a half-lives of about 210 thousand years and about 1.6 million years, respectively. Technetium-99 is water-soluble and there is fear that it is, after a long period of time, eluted and released out of a barrier in the form of ions of TcO.sub.4.sup.-, etc. following geologic disposal. With respect to iodine-99, considering its safety from the viewpoint of migration into the ground water when the iodine-99 is solidified and then subject to geologic disposal, there are the problems that this material has a high solubility and has a low absorption by the barrier. Thus, from a viewpoint of reduction of the environmental load, these long-lived FP nuclides should preferably be transformed into the other stable nuclides by a suitable method before a final disposal thereof. For example, the technetium-99 and iodine-129 are transformable, by a neutron absorption reaction, into non-radioactive and stable ruthenium (Ru) and xenon (Xe). Technetium-99 and iodine-129, which are long-lived FP nuclides, have larger neutron absorption cross sections for thermal neutron energy than for fast neutron energy, and have larger resonance absorption regions in a lower energy region (approximately 5 eV, etc.). Therefore, in order to transmutate these long-lived FP nuclides by neutron absorption reactions, it is advisable to slow down to some extent fast neutrons which have been generated by nuclear fission, and then use the same as resonance energy neutrons or thermal neutrons. Both thermal neutron reactors and fast neutron reactors are considered as reactors in which to transmute long-lived FP nuclides. A thermal neutron reactor utilizes moderated or slowed-down neutrons and, therefore, the nuclides can be transmuted, to some extent, by loading pins which contain technetium-99 or iodine-129 into the reactor core. A fast neutron reactor, on the other hand, requires the fast neutrons to be moderated and, therefore, has utilized an assembly having a structure in which long-lived FP nuclides-containing pins and moderator-containing pins are housed together in a wrapper tube, for the purpose of transmutation of these long-lived FP nuclides. A conventional assembly for the purpose of transmutation is shown in FIG. 6, in which moderator-containing pins 12 and FP-containing pins 14 are disposed in a dispersed arrangement in a wrapper tube 10 as illustrated. The moderator pin 12 is composed of a cladding tube which contains moderator material and nothing else, and the FP pin 14 is composed of a cladding tube which contains the material including long-lived FP nuclides, and nothing else. In the case where the assemblies for the transmutation purposes as described above are loaded into a blanket region of the fast reactor, the transmutation rate of the long-lived FP nuclides is approximately 2.0-2.5% which is not as much as the requirement. The inventor of the present invention considers that the reason for such a low transmutation rate resides in a remarkably high self-shielding effect of the neutrons in the FP pins such that the nuclear reactions between the FP and the neutrons are carried out only near the surface of the FP pins, with the result that the neutrons are prevented from entering deep into the FP pins. This problem would be solved to a limited extent by an attempt at forming thinner FP pins but this measure would raise other serious problems in that the necessary number of pins is increased, resulting in deficiencies in production, working effect and cost performance. Further, in the conventional structure of the transmutation-purpose assemblies as described above, the arrangement of the FP pins and the moderator pins in the space of the assembly is complex, resulting in less working efficiency and difficulties in inspection of the pins. SUMMARY OF THE INVENTION Accordingly, the present invention is proposed to solve the above problem, and it is therefore an object of the present invention to provide an improved assembly for transmutation which permits an efficient transmutation (i.e., transformation into stable nuclides by a nuclear transformation), in a nuclear reactor, of a long-lived radioactive material (especially, long-lived FP nuclides such as technetium-99 or iodine-129, etc.) which was produced in a nuclear reactor. Another object of the invention is to provide an assembly for transmutation of a long-lived radioactive material which permits a high workability of production and a high operability of inspection and also meets with a requirement for reduction of costs. In a first aspect of the present invention, there is provided an assembly for transmutation of a long-lived radioactive material comprising: wire-type members of a long-lived radioactive material comprised of metals, alloys or compounds including long-lived fission product nuclides, PA1 a moderator material surrounding the wire-type members, PA1 cladding tubes each containing therein the wire-type members surrounded by the modulator material to thereby form FP (fission products) pins, and PA1 a wrapper tube housing therein the FP pins and nothing else. PA1 thin ring-type (or tubular) members of a long-lived radioactive material comprised of metals, alloys or compounds including long-lived fission product nuclides, PA1 a moderator material surrounding an inner surface and an outer surface of the thin ring-type members, PA1 cladding tubes each containing therein the thin ring-type members surrounded by the moderator material to thereby form FP pins, and PA1 a wrapper tube housing therein the FP pins and nothing else. PA1 wire-type members of a long-lived radioactive material comprised of metals, alloys or compounds including long-lived fission product nuclides, PA1 a moderator material surrounding the wire-type members, PA1 cladding tubes each containing therein the wire-type members surrounded by the moderator material to thereby form FP pins, and PA1 a wrapper tube housing therein the FP pins solely. PA1 thin ring-type members of a long-lived radioactive material comprised of metals, alloys or compounds including long-lived fission product nuclides, PA1 a moderator material surrounding an inner surface and an outer surface of the thin ring-type members, PA1 cladding tubes each containing therein the thin ring-type members surrounded by the moderator material to thereby form FP pins, and PA1 a wrapper tube housing therein the FP pins solely. In the transmutation assembly described above, the wire-type members of the long-lived radioactive material may be located in a dispersed state and each of the dispersed wire-type members is surrounded by the moderator material. In a second aspect of the invention, there is provided an assembly for transmutation for a long-lived radioactive material comprising: In a third aspect of the invention, there is provided a reactor core for a fast reactor comprising a core region, a blanket region and a shield region, wherein transmutation assemblies are selectively and at least partly loaded into the core region, the blank region or the shield region, each of said transmutation assemblies comprising: In a fourth aspect of the invention, there is provided a reactor core for a fast reactor comprising a core region, a blanket region and a shield region, wherein transmutation assemblies are selectively and at least partly loaded into the core region, the blanket region or the shield region, each of said transmutation assemblies comprising: In the present invention, wire-type members of a long-lived radioactive material are surrounded by a moderator material and then the wire-type members surrounded by the moderator material are installed in cladding tubes to form FP pins and the FP pins, and nothing else, are housed in a wrapper tube. The long-lived radioactive material is comprised of metals, alloys or compounds including long-lived fission product nuclides. The wire-type members described above can be placed in a dispersion arrangement with the enclosure by the moderator material. In another embodiment of the invention, the wire-type members described above can be replaced by a single or a plurality of ring-type or thin-walled tubular member(s), and the thin ring-type member(s) is or are surrounded by the moderator material on the outer and inner surfaces thereof and then placed into the cladding tubes to thereby form the FP pins. Only the FP pins are housed in the wrapper tube. In order to carry out a transmutation treatment, the transmutation assemblies of the long-lived radioactive material as described above can be loaded selectively and at least partly into a core region, a blanket region or a shield region of a reactor core of a fast reactor. From a viewpoint of reducing possible influence on the reactor core characteristics, it is optimal to load the transmutation assemblies into the blanket region in the reactor core.
	claims	1. A charged particle irradiation system comprising:an accelerator for extracting a charged particle beam;charged-particle-beam scanning equipment; anda charged-particle-beam-position monitor system,wherein said charged particle irradiation system further includes a control unit for calculating a beam position at a target position on the basis of a signal received from the charged-particle-beam-position monitor system to control the charged-particle-beam scanning equipment so that the charged particle beam is moved to a desired irradiation position at the target position, andwherein in each predetermined cycle of the accelerator, said control unit corrects the value of an excitation current applied to the charged-particle-beam scanning equipment on a specified cycle basis on the basis of information about the position and the angle of the charged particle beam. 2. The charged particle irradiation system according to claim 1, wherein:said charged particle irradiation system is equipped with two of the charged-particle-beam-position monitor systems; andwherein in each predetermined cycle of the accelerator, said control unit corrects the value of the excitation current applied to the charged-particle-beam scanning equipment on a specified cycle basis on the basis of information about the position and the angle of the charged particle beam which are obtained from a first charged-particle-beam position detected by a first charged-particle-beam-position monitor system and a second charged-particle-beam position detected by a second charged-particle-beam-position monitor system. 3. The charged particle irradiation system according to claim 2, wherein:in each extraction cycle of the accelerator, said control unit corrects the value of an excitation current applied to the charged-particle-beam scanning equipment on the basis of information about the position and the angle of the charged particle beam. 4. The charged particle irradiation system according to claim 2, wherein:said control unit sets an excitation current value for the charged-particle-beam scanning equipment such that in the next cycle, a point on the opposite side of the charged particle beam position with respect to the desired irradiation position, which point is away from the beam position by the distance between the beam position and the desired position at the target position, is irradiated with such a charged particle beam with which the beam position agrees with the desired irradiation position. 5. The charged particle irradiation system according to claim 2, wherein:even when the charged-particle-beam scanning equipment scans the charged particle beam, said control unit calculates the distance between the desired position and the charged particle beam position at the target position on the basis of the distance between the position of the charged particle beam traveling toward the desired irradiation position at the position of the charged-particle-beam-position monitor system and the path of the charged particle beam. 6. The charged particle irradiation system according to claim 2, wherein:said first charged-particle-beam-position monitor system is disposed upstream of the charged-particle-beam scanning equipment and said second charged-particle-beam-position monitor system is disposed downstream of the charged-particle-beam scanning equipment. 7. The charged particle irradiation system according to claim 2, wherein:said first and second charged-particle-beam-position monitor systems are disposed downstream of the charged-particle-beam scanning equipment. 8. The charged particle irradiation system according to claim 2, wherein:said first and second charged-particle-beam-position monitor systems are disposed upstream of the charged-particle-beam scanning equipment. 9. The charged particle irradiation system according to claim 3, wherein:said control unit identifies the progress of the phase of the extraction cycle from a radiated charge amount in the same extraction cycle. 10. The charged particle irradiation system according to claim 3, wherein:said control unit identifies the progress of the phase of the extraction cycle from a numerical value that is obtained by dividing a radiated charge amount in the same extraction cycle by an accumulated charge amount. 11. The charged particle irradiation system according to claim 3, wherein:said control unit corrects the charged particle beam irradiation position in the next extraction cycle by a function in which a position to which the charged particle beam irradiation position changes according to the phase in the extraction cycle is approximated by a phase function of the extraction cycle. 12. The charged particle irradiation system according to claim 1, wherein:said charged particle irradiation system is equipped with one charged-particle-beam-position monitor system, said monitor system being capable of measuring both the position and the angle of the charged particle beam. 13. The charged particle irradiation system according to claim 1, wherein:if the difference between the charged particle beam position and the desired irradiation position at the target position exceeds a threshold value, a safety interlock system is operated, said charged particle beam position having been calculated by the control unit. 14. The charged particle irradiation system according to claim 12, wherein:in each extraction cycle of the accelerator, said control unit corrects the value of an excitation current applied to the charged-particle-beam scanning equipment on the basis of information about the position and the angle of the charged particle beam. 15. The charged particle irradiation system according to claim 12, wherein:said control unit sets an excitation current value for the charged-particle-beam scanning equipment such that in the next cycle, a point on the opposite side of the charged particle beam position with respect to the desired irradiation position, which point is away from the beam position by the distance between the beam position and the desired position at the target position, is irradiated with such a charged particle beam with which the beam position agrees with the desired irradiation position. 16. The charged particle irradiation system according to claim 12, wherein:even when the charged-particle-beam scanning equipment scans the charged particle beam, said control unit calculates the distance between the desired position and the charged particle beam position at the target position on the basis of the distance between the position of the charged particle beam traveling toward the desired irradiation position at the position of the charged-particle-beam-position monitor system and the path of the charged particle beam. 17. The charged particle irradiation system according to claim 14, wherein:said control unit identifies the progress of the phase of the extraction cycle from a radiated charge amount in the same extraction cycle. 18. The charged particle irradiation system according to claim 14, wherein:said control unit identifies the progress of the phase of the extraction cycle from a numerical value that is obtained by dividing a radiated charge amount in the same extraction cycle by an accumulated charge amount. 19. The charged particle irradiation system according to claim 14, wherein:said control unit corrects the charged particle beam irradiation position in the next extraction cycle by a function in which a position to which the charged particle beam irradiation position changes according to the phase in the extraction cycle is approximated by a phase function of the extraction cycle.
	claims	1. A radiation emitting device comprising:a radiation emitter configured for emitting X-rays;an emitter switch, the radiation emitter being secured to the emitter switch;a disk-shaped collimator, the collimator having a central axial through-hole portion and a plurality of radial apertures configured for collimating the X-rays emitted from the radiation emitter into pencil beams, the through-hole portion receiving the radiation emitter and the emitter switch therein;a rotating mechanism coupled to the through-hole portion of the collimator for rotating the collimator;an annular shielding enclosure having an opening configured for allowing the pencil beams to exit therethrough, wherein the shielding enclosure encloses the collimator, the radiation emitter and the emitter switch therein; whereinthe radiation emitter is jointly and axially movable with respect to the emitter switch in the through-hole portion between an off position and an on position, wherein in the off position, the radiation emitter is misaligned with any one of the radial apertures thus ensuring the X-rays emitted from the radiation emitter blocked by the shielding enclosure, and wherein in the on position, the radiation emitter is aligned with one of the radial apertures therefore ensuring the X-rays emitted from the radiation emitter exit from the opening of the shielding enclosure. 2. The radiation emitting device, as recited in claim 1, wherein the radiation emitter is radially engaged with the collimator such that the radiation emitter, the emitter switch and the collimator are collectively rotatable relative to the shielding enclosure. 3. The radiation emitting device as recited in claim 1, wherein in an on position, the radiation emitter and the emitter switch are at rest relative to the shielding enclosure. 4. The radiation source device as recited in claim 1, further comprising a frame movable along a predetermined direction, the shielding enclosure being mounted on the frame.
	claims	1. A lithographic apparatus comprising a filter device, the filter device comprising a plurality of foils attached to a holder able to rotate around a rotation axis, the foils being arranged substantially parallel to the rotation axis and comprising a uni-directional carbon-fiber composite material selected from the group consisting of a carbon-carbon composite and a carbon-silicon carbide composite. 2. The lithographic apparatus of claim 1, wherein the direction of the fibers in the carbon-fiber composite material is transverse to the rotation axis. 3. The lithographic apparatus of claim 1, wherein the foils comprise one layer of uni-directional carbon-fiber composite material. 4. The lithographic apparatus of claim 1, wherein the foils comprise a first layer of uni-directional carbon-fiber composite material and a second layer of uni-directional carbon-fiber composite material, wherein the direction of the fibers in the first layer is transverse to the direction of the fibers in the second layer. 5. The lithographic apparatus of claim 1, wherein the foils comprise a first layer of uni-directional carbon-fiber composite material and a second layer of uni-directional carbon-fiber composite material, wherein the direction of the fibers in the first layer have a first direction angle relative to a normal perpendicular to the rotation axis and wherein the direction of the fibers in the second layer have a second direction angle relative to the normal, and wherein the first and the second direction angles are in the range of 0°-10° and wherein the mutual angle between the direction of the fibers in the first layer and the second layer is larger than 0° and equal to or smaller than 10°. 6. The lithographic apparatus of claim 1, wherein the foils comprise 2-5 layers of composite material. 7. The lithographic apparatus of claim 1, wherein the foils have a maximum foil thickness in the range of 0.05- 1.2 mm. 8. The lithographic apparatus of claim 1, wherein the foils have a maximum foil thickness in the range of 0.1- 0.4 mm. 9. The lithographic apparatus of claim 1, wherein the holder comprises a plurality of sleeves, wherein each foil comprises a tail part, wherein the sleeves are constructed to receive the tail parts and to prevent release of the foils from the holder in a direction perpendicular to the rotation axis. 10. The lithographic apparatus of claim 1, wherein the filter device comprises 50-200 foils. 11. The lithographic apparatus of claim 1, wherein at least part of the holder comprise a carbon-fiber composite material selected from the group consisting of a carbon-carbon composite and carbon-silicon carbide composite. 12. The lithographic apparatus of claim 1, further comprising a Sn plasma source of radiation constructed to generate EUV radiation. 13. A lithographic apparatus comprising a filter device, the filter device comprising a plurality of foils attached to a holder which is able to rotate around a rotation axis, the foils being arranged parallel to the rotation axis and comprising a uni-directional carbon-fiber composite material which does not substantially react with liquid Sn at a temperature of at least 1000° C. 14. A filter device comprising a plurality of foils attached to a holder which is able to rotate around a rotation axis, the foils being arranged substantially parallel to the rotation axis and comprising a uni-directional carbon-fiber composite material selected from the group consisting of carbon-carbon composite and carbon-silicon carbide composite. 15. The filter device of claim 14, wherein at least part of the holder comprise a carbon-fiber composite material selected from the group consisting of carbon-carbon composite and carbon-silicon carbide composite. 16. A method for the production of a foil for a filter device, the method comprising:a. providing a resin containing pre-impregnated sheet;b. curing the resin;c. optionally reducing the thickness of at least part of the product obtained at b);d. carbonizing the product obtained at b) or c);e. optionally performing one or more times a densifying process, wherein the densifying process comprises infiltrating the carbonized product with a carbon-containing compound and subsequently carbonizing the infiltrated product;f. graphitizing the product obtained at d) or e);g. optionally reducing the thickness of at least part of the product obtained at f)wherein the method comprises reducing the thickness by process c), or by process g), or by both process c) and g), and wherein the foil comprises a uni-directional carbon-fiber composite material selected from the group consisting of carbon-carbon composite and carbon-silicon carbide composite. 17. The method according to claim 16, wherein the resin containing pre-impregnated sheet comprises a laminate of resin containing pre-impregnated sheets. 18. The method according to claim 16, further comprising, after the process b) and before process d), performing one or more times a laminating process, wherein the laminating process comprises:a1. arranging a further resin containing pre-impregnated sheet to the product obtained at b) or c) to obtain a laminate of the product obtained at b) or c) and the further resin containing pre-impregnated sheet; andb1. curing the resin. 19. The method according to claim 18, further comprising reducing the thickness of at least part of the product obtained at b1. 20. The method according to claim 16, comprising performing process e). 21. A device manufacturing method, comprising:patterning a beam of radiation;projecting the patterned beam of radiation onto a target portion of a substrate; andfiltering the beam of radiation using a filter device, the filter device comprising a plurality of foils attached to a holder able to rotate around a rotation axis, the foils being arranged substantially parallel to the rotation axis and comprising a uni-directional carbon-fiber composite material selected from the group consisting of a carbon-carbon composite and a carbon-silicon carbide composite. 22. The device manufacturing method of claim 21, wherein during lithographic processing, rotating the filter device about the rotation axis.
	summary
	summary
	abstract	A fission chamber count rate measurement device and to the associated fission chamber calibration device; the count rate measurement device comprises: (1) a measurement cell, which contains the fission chamber (CH); (2) a neutron generator, which emits neutrons in the form of periodic pulses towards the fission chamber; (3) a neutron counter (K), which detects and counts the neutrons emitted by the neutron generator; and (4) a computing circuit, which delivers, over a predetermined time interval, a fission chamber count rate normalized with reference to the number of neutrons counted by the neutron counter (K).
043549993	claims	1. An improved plasma confining apparatus of the type having electromagnetic Ioffe bar windings, wherein the improvement comprises a plurality of interconnected Ioffe bar windings each winding being would in a spherical configuration about a plurality of different transverse axes of said spherical configuration, said axes being offset from the center of the spherical configuration, said spherical configuration comprising two adjacent concentric spherical surfaces, each said spherical surface having different finite radii, said Ioffe bar windings lying between said two adjacent concentric spherical surfaces, each said winding being composed of arcuate portions, said Ioffe bar windings forming a closed curve, at least one of said portions extending in said spherical configuration for a distance substantially one half the circumference of said spherical configuration. 2. An improved plasma containing and confining apparatus of the type defining an enclosed, evacuated reaction chamber and having surrounding electromagnetic confinement windings, wherein the improvement comprises said reaction chamber being generally spherical combined with a plurality of interconnected Ioffe bar windings each winding being would outside said reaction chamber in a spherical configuration about a plurality of different transverse axes of said reaction chamber, said axes being offset from the center of the spherical configuration, said spherical configuration comprising two adjacent concentric spherical surfaces, each said spherical surface having different finite radii, said Ioffe bar windings lying between said two adjacent concentric spherical surfaces, each said winding being composed of arcuate portions, said Ioffe bar windings forming a closed curve, at least one of said portions extending in said spherical configuration for a distance substantially one half the circumference of said spherical configuration. 3. An apparatus in accordance with claim 2 wherein there is further provided a source of alternating current connected to energize said windings, said source providing voltage alternations at the natural cyclotron resonance frequency of said chamber.
	claims	1. A method for suppressing corrosion of a carbon steel member, comprising the steps of:forming a nickel metal film on a surface of the carbon steel member in a nuclear power plant;forming a nickel ferrite film on a surface of the nickel metal film; andafter the formation of the nickel ferrite film, transforming the nickel metal film into a nickel ferrite film. 2. The method for suppressing corrosion of the carbon steel member according to claim 1, wherein the step of forming the nickel metal film is performed by exposing the surface of the carbon steel member to a film forming liquid containing nickel ions, whose pH is adjusted to a value in a range between 4.0 and 9.0. 3. The method for suppressing corrosion of the carbon steel member according to claim 1, wherein the step of forming the nickel metal film is performed by exposing the surface of the carbon steel member to a film forming liquid containing nickel ions and iron (II) ions, whose pH is adjusted to a value in a range between 4.0 and 9.0. 4. The method for suppressing corrosion of the carbon steel member according to claim 2, wherein the film forming liquid contains formic acid. 5. The method for suppressing corrosion of the carbon steel member according to claim 1, wherein the step of forming the nickel ferrite film is performed by exposing the surface of the nickel metal film to a film forming liquid including a first agent containing iron (II) ions, a second agent containing nickel ions, and a third agent for oxidizing the iron (II) ions, whose pH is adjusted by addition of a pH adjustment agent to a value within a range between 5.5 and 9.0. 6. The method for suppressing corrosion of the carbon steel member according to claim 1, wherein the steps of forming the nickel metal film and forming the nickel ferrite film on the surface of the nickel metal film are performed in a period when operation of the plant is being shut down. 7. The method for suppressing corrosion of the carbon steel member according to claim 6, wherein the step of transforming the nickel metal film into the nickel ferrite film is performed in the period when the operation of the plant is being shut down. 8. The method for suppressing corrosion of the carbon steel member according to claim 1, wherein the step of transforming the nickel metal film into the nickel ferrite film is performed by exposing a surface of the nickel ferrite film formed on the nickel metal film to water containing dissolved-oxygen at 150° C. or above. 9. A suppression method for suppressing corrosion of a carbon steel member, comprising the steps of:forming a nickel metal film on a surface of the carbon steel member in a nuclear power plant;forming a nickel ferrite film on a surface of the nickel metal film; andafter the formation of the nickel ferrite film, transforming the nickel metal film into a nickel ferrite film,wherein the steps of forming the nickel metal film and forming the nickel ferrite film on the surface of the nickel metal film are performed in a period when operation of the plant is being shut down, and the step of transforming the nickel metal film into the nickel ferrite film is performed when the plant is in operation. 10. The method for suppressing corrosion of the carbon steel member according to claim 9, wherein the step of forming the nickel metal film is performed by exposing the surface of the carbon steel member to a film forming liquid containing nickel ions, whose pH is adjusted to a value in a range between 4.0 and 9.0. 11. The method for suppressing corrosion of the carbon steel member according to claim 9, wherein the step of forming the nickel metal film is performed by exposing the surface of the carbon steel member to a film forming liquid containing nickel ions and iron (II) ions, whose pH is adjusted to a value in a range between 4.0 and 9.0. 12. The method for suppressing corrosion of the carbon steel member according to claim 10, wherein the film forming liquid contains formic acid. 13. The method for suppressing corrosion of the carbon steel member according to claim 9, wherein the step of forming the nickel ferrite film is performed by exposing the surface of the nickel metal film to a film forming liquid including a first agent containing iron (II) ions, a second agent containing nickel ions, and a third agent for oxidizing the iron (II) ions, whose pH is adjusted by addition of a pH adjustment agent to a value within a range between 5.5 and 9.0. 14. The method for suppressing corrosion of the carbon steel member according to claim 9, wherein a temperature of the film forming liquid is adjusted between 60 and 100° C. 15. The method for suppressing corrosion of the carbon steel member according to claim 9, wherein the step of transforming the nickel metal film into the nickel ferrite film is performed by exposing a surface of the nickel ferrite film formed on the nickel metal film to water containing dissolved-oxygen at 150° C. or above.
	summary
	summary
	claims	1. A transportable container for nuclear fuel comprising:an outer container bounding an interior and defining an overall volume;a thermal insulation material disposed within the interior bounded by the outer container, the thermal insulation material bounding an internal cavity;a plurality of sleeves disposed within the cavity; andone or more fuel containers received within at least one of the sleeves, each of the one or more fuel containers having an internal volume and a releasable lid. 2. A container according to claim 1 in which the sleeves are surrounded by a neutron absorbing material, the neutron absorbing material filling the internal cavity apart from the inside of the sleeves. 3. A container according to claim 1 in which the volume of the internal cavity outside of the sleeves is filled by neutron absorbing material or neutron absorbing material which incorporates lower density materials. 4. A container according to claim 1 in which the sleeves are rigidly separated from one another. 5. A container according to claim 1 in which the outer container is comprised of steel, the sleeves are comprised of stainless steel and have a substantially circular transverse cross-section, the sleeves having an internal diameter that is substantially equal to an external diameter of the fuel containers, the sleeves being rigidly separated from one another, the sleeves being surrounded around their entire circumference by a neutron absorbing material, the fuel containers being comprised of stainless steel having a substantially cylindrical configuration, and nuclear fuel being disposed within the fuel containers in plastic bags. 6. A container according to claim 1 in which the outer container is provided with a lid. 7. A container according to claim 1 in which only one fuel container is provided in each sleeve. 8. A transportable container for nuclear fuel comprising:an outer container bounding an interior and defining an overall volume;a thermal insulation material disposed within the interior bounded by the outer container, the thermal insulation material comprising one or more base layers and one or more wall layers; anda plurality of chambers being provided within bounds defined by the thermal insulation material, one or more fuel containers being provided within each of a plurality of the chambers. 9. A container according to claim 8 in which the chambers are surrounded by a neutron absorbing material, the neutron absorbing material filling the bounds defined by the insulation apart from the inside of the chambers. 10. A container according to claim 8 in which neutron absorbing material or neutron absorbing material which incorporates lower density materials fills the volume around the sleeves. 11. A container according to claim 8 in which the internal bounds of the thermal insulation material contact a neutron absorbing material. 12. A container according to claim 8 in which the internal insulation is neutron absorbing. 13. A container according to claim 12 in which the interior bounds of the neutron absorbing insulation contact a neutron absorbing material. 14. A container according to claim 12 in which the neutron absorbing material is loaded with boron. 15. A container according to claim 8 in which the outer container is comprised of steel, the sleeves are comprised of stainless steel and have a substantially circular transverse cross-section, the sleeves having an internal diameter that is substantially equal to an external diameter of the fuel containers, the sleeves being rigidly separated from one another, the sleeves being surrounded around their entire circumference by a neutron absorbing material, the fuel containers being comprised of stainless steel having a substantially cylindrical configuration, and nuclear fuel being disposed within the fuel containers in plastic bags. 16. A transportable container for nuclear fuel, the container comprising:an outer container, the outer container being provided with a thermal insulation material disposed therein;a plurality of laterally spaced apart sleeves provided within the outer container; andone or more fuel containers received within the one or more of the sleeves, the fuel containers each being provided with a releasable lid for the fuel container. 17. A container according to claim 16 in which the releasable lid for the fuel container seals the fuel container when fastened and the releasable lid for the outer container seals the outer container when fastened. 18. A container according to claim 16 in which the outer container and outer container lid provides a first barrier and the fuel container and fuel container lid provides a second barrier between the nuclear fuel and the exterior of the outer container. 19. A transportable container for nuclear fuel comprising:an outer container bounding an interior and defining an overall volume;a thermal insulation material disposed within the interior bounded by the outer container, the thermal insulation material bounding an internal cavity;four or more sleeves disposed within the internal cavity; andone or more fuel containers received within at least one of the sleeves, the sleeves having an internal diameter that is substantially equal to an external diameter of the fuel containers, the sleeves being rigidly separated from one another. 20. A transportable container according to claim 19, in which the sleeves have a substantially circular transverse cross-section and eight or nine sleeves are provided. 21. A transportable container for nuclear fuel comprising:an outer container bounding an interior and defining an overall volume;a thermal insulation material disposed within the interior bounded by the outer container, the thermal insulation material bounding an internal cavity;a plurality of sleeves disposed within the cavity; andwherein the nuclear fuel is disposed within one or more plastics bags; andwherein the one or more plastics bags are disposed within one or more fuel containers, each of the fuel containers being provided with a releasable lid; andwherein the one or more fuel container are each received within one of the plurality of sleeves. 22. A transportable container according to claim 21, in which eight or nine sleeves are provided and at least eight of the sleeves receive a fuel container. 23. A transportable container for nuclear fuel comprising:an outer container bounding an interior and defining an overall volume;a thermal insulation material disposed within the interior bounded by the outer container, the thermal insulation material bounding an internal cavity;a plurality of sleeves disposed within the cavity; andone or more fuel containers received within at least one of the sleeves, each of the one or more fuel containers having an internal volume and a releasable lid; andwherein the nuclear fuel is in the form of powder or pellets and wherein the nuclear fuel is to subsequently be used in fuel rod manufacture.
	abstract	A projection optical unit images an object field in an image field. The projection optical unit includes a plurality of mirrors guides imaging light from the object field to the image field. At least two of the mirrors are arranged directly behind one another in the beam path of the imaging light for grazing incidence with an angle of incidence of the imaging light which is greater than 60°. This results in an imaging optical unit that can exhibit a well-corrected imageable field with, at the same time, a high imaging light throughput.
048851221	summary	RELATED APPLICATIONS This application is related by subject matter to copending application Ser. No. 925,861 which is assigned to the assignee of the present invention and which is incorporated herein by reference. BACKGROUND OF THE INVENTION This invention relates to clamps for sealing the instrumentation ports associated with nuclear reactor systems. More particularly, the present invention relates to high quality clamps for maintaining a proper seal at the interface between reactor vessel head penetrations and the thermocouple instrument columns. Due to the risks associated with operating a nuclear power plant, the design and quality standards associated with nuclear reactor equipment are extremely high and stringent. Accordingly, problems which are capable of straight forward solution in a non nuclear environment are difficult and demanding in the context of a nuclear reactor facility. For example, it is generally required in many industrial settings to monitor the pressure, temperature, and other parameters of various operating equipment. In the environment of a nuclear power plant, this can be extremely dangerous due to the potential for escape of radioactive materials. Accordingly, it is imperative in these situations that the instrumentation used to make such measurements be precisely designed to prevent such leaks. While the possible escape of radioactive material from a nuclear power plant is minimized by the containment building surrounding the nuclear steam supply system, the working conditions inside the containment building are potentially hazardous even during normal plant operation. This is particularly true during refueling of the reactor when high background and high airborne particulate radioactivity exists in the containment building. Safety regulations set maximum dose limits for the presence of workers in these locations during plant operation and refueling. Many of these same locations are also hazardous during refueling due to the high ambient temperatures. In many situations, these locations are also not easily accessible and a safe work platform is not available. It is accordingly desirable to provide clamping apparatus which are quickly and easily assembled so as to minimize worker exposure to such hazardous conditions. Such quick and efficient repair and/or replacement of instrument port clamps is also highly desirable from the economic point of view since it minimizes the down time of the nuclear plant and hence the cost of providing replacement electricity. In order to more clearly understand the present invention, one typical instrumentation port interface is revealed in FIG. 1. As revealed by this illustration, a lower conduit or flange 10 is coupled or otherwise mounted to the vessel (not shown) whose parameter is to be measured while the upper conduit or flange 11 is coupled or otherwise mounted to the parameter measuring assembly (not shown). In the particular application of a nuclear power plant, the lower portion of flange 10 is threaded and welded onto the head penetration. Flanges 10 and 11 are generally tubular in shape and have upper and lower surfaces respectively which are designed to engage one another in a sealing manner with respect to gasket 12. In order to effectively compress gasket 12 and seal the interface between flanges 10 and 11, it is desirable for a clamping apparatus to exert axially pressure on flanges 10 and 11. For the instrument port interface shown in FIG. 1, the pressure exerted on each flange should be directed towards the interfacing end of that flange. That is, the clamping apparatus should exert an upward axial force on flange 10 while exerting a substantially equal and axially downward force on flange 11. In this way, the interface between flanges 10 and 11 is properly sealed by gasket 12. The seal between flange 10 and flange 11 is an important safety consideration in the design of a nuclear power plant reactor. It will be appreciated by those skilled in the art that such flange interfaces are generally located in regions of the plant having a high radioactivity level and high process temperatures. Because of these special circumstances, high quality clamps capable of sealing the interface between flanges 10 and 11 are not only desirable but necessary. In some applications, it is desirable to construct such clamps from high strength material. In addition, it is highly critical to worker safety that the clamping apparatus used to seal such interface be quickly and easily installed and removed. One heretofore used clamping apparatus, generally designated as 20, is shown in FIG. 2. The clamping apparatus 20 consists of three essentially identical body members 13A, 13B, and 13C. Each body member spans an arc of approximately 110.degree.. An interbody gap 15 of about 10.degree. exists between the body members. Each end of the body members 13A, 13B and 13C contains a flanged portion which is used to attach the body members together. A cap screw 14 (as shown) or other holding means is passed through the flanged ends and holds the body members in a generally. ring-shape while the clamp is assembled on flanges 10 and 11. The use of the clamp 20 on a nuclear reactor vessel instrumentation port interface as shown in FIG. 1 will now be described. Due to its configuration and weight, the clamp 20 of FIG. 2 is generally brought to the instrument port interface in disassembled form. At least two workers are then generally required to assembly clamp 20 in situ around the outer portion of the interface between flanges 10 and 11. Workmen only have access to flanges 10 and 11 from radially outside the reactor vessel head because of the cooling shroud and other equipment permanently installed thereabove. The specified procedure for operation of the heretofore used clamping apparatus requires the use of an axial loading device which seats the gasket prior to the application of the clamp. Such axial loading devices are generally cumbersome and heavy, making the installation thereof extremely difficult. The requirement of this axial loading device also restricts the work space available and therefore complicates the assembly of clamp 20. Once the axial loading device is properly positioned, the interbody gaps 15 must be carefully adjusted so as to be substantially equivalent in order to achieve generally uniform contact and pressure on the flanges 10 and 11, and to minimize cap screw shank bending. The cap screws 14 are generally torqued to about 100 ft/lb Torquing to this extent may require relatively long torquing systems. It should be noted that, in many applications, over torquing of the cap screws 14 may result in overcompression of gasket 12. For many gaskets, overcompression has a serious detrimental impact on the sealing capacity of the gasket. Prior art clamping apparatus generally used space limiters between the flanges in order to prevent such overcompression of the gasket. It is apparent from the above description that the procedures and apparatus required for the assembly of clamp 20 and other prior clamping devices are thus time consuming and present a large potential for improper installation. The above disadvantages are even more pronounced when it is considered that such a clamp must be installed in awkward and precarious positions requiring workers to be tethered by ropes and/or other safety gear and that workers are required to wear cumbersome gear such as masks, heavy gloves, and radiation suits with respirators. While the use of articulated clamps to overcome some of the disadvantages described above has been known, the heretofore articulated clamps did not completely overcome the problems and difficulties associated with use of such clamps in a nuclear power plant environment. For example, the heretofore used articulated clamps did not solve the serious problem of possible gasket compression and as a result required the use of space limiters. While such space limiters may have been adequate for the intended purpose, their use is prevented in many nuclear power plants systems due to existing space restrictions and existing instrument port flange configurations. SUMMARY OF THE INVENTION It is an object of the present invention to provide clamps and clamping systems for use in nuclear reactor and nuclear power plant environments, said clamps providing quick, safe, and easy installation in such environments. It is a further object of the present invention to provide clamps which apply uniform circumferential clamping pressure without the aid of axial loading devices. It is yet another object of the present invention to provide a clamping device which avoids overcompression of the sealing gasket without the need for a space limiter. It is another object of the present invention to provide clamps which minimize the amount of time workers are exposed to hazardous working environments, and to minimize the number of workers so exposed. It is a still further object of the present invention to provide clamps which are installed and removed by application of relatively low torque without the need for special tools. According to one embodiment of the present invention, the above and other objects are satisfied by a clamp comprising: a plurality of intermediate body members pivotally joined together to form an intermediate body portion having two ends; two end body members, each of said end body members having an unflanged end pivotally attached to one end of said intermediate body portion and an flanged end; and means for joining the flanged ends. An instrument port clamp according to another embodiment of the present invention comprises: two end body members, each of said body members having a first end and a second end; means for linking said first ends; and means for releasably joining said second ends, the spacing between said second ends being adjustable when said joining means is released. Another embodiment provides clamping systems for sealing the interface between two generally tubular bodies comprising: each of said tubular bodies having a clamp receiving portion thereof; and at least one generally ringshape clamp in engagement with said clamp receiving portions, said clamp comprising: two end body members, each of said body members having a first end and a second end; means for linking said first ends such that the spacing between said second ends is adjustable; and means for joining said second ends.
	description	The present invention generally refers to an irradiation system, and in particular to such a system being able to direct a radiation beam into multiple treatment rooms. During the past decades there have been considerable developments within the fields of radiation therapy and diagnosis. The performance of external beam radiation therapy accelerators, brachytherapy and other specialized radiation therapy equipment has improved rapidly. Developments taking place in the quality and adaptability of therapeutic radiation beams have included new targets and filters, improved accelerators, increased flexibility in beam-shaping through new applicators, collimator and scanning systems and beam compensation techniques. Also improved dosimetric and geometric treatment verification methods have been introduced. Furthermore, new treatment planing systems capable of biological optimization of the intensity distribution of the delivered beams are now being available. In the field of multiple and single fraction radiation therapy and diagnostic imaging, a common general method is to position the patient on a couch. A radiation head and gantry are directing a diagnostic or therapeutic beam onto the patient in order to deliver radiation to a certain target or treatment volume, e.g. a tumor. Such a typical radiation machine according to the prior art is schematically illustrated in FIG. 1. The radiation machine comprises an isocentric gantry 80 designed in a general L shape and a rotational support provided at one axial end of the body of the machine for supporting the gantry 80. Thus, the gantry 80 can rotate around a rotation axis 30 relative the support in order to deliver a radiation beam, schematically illustrated by 10, from a radiation head 20 into a target volume 55 of a patient 50 positioned on a patient couch 40. Most of the radiation therapy machines of today, including the machine in FIG. 1, comprise an isocentric gantry design. In such a design, the tissue or target volume 55 to be radiated is preferably positioned around a so-called isocenter typically formed by the intersection of three axes at a common point. These axes include the gantry rotation axis 30, the central axis of the radiation beam 10, the major rotational axis 45 of the treatment couch 45, which is also the rotation axis of the collimator head 20 in the figure. A problem with such prior art radiation therapy machines is their limited capacity in terms of the total number of patients that can be treated in a given time interval. Although the actual irradiation is rather quick, i.e. it typically lasts a few minutes (1-2.5 minutes), a much longer treatment set-up normally precedes the irradiation. During such a set-up, the personnel positions the patient to be treated as accurately as possible, typically based on a treatment plan, which has been developed or compiled earlier based on diagnostic data, radiation beam data, etc. After placing the patient on the couch, but before the actual radiation therapy treatment, a treatment set-up is typically performed to test and verify the beam directions and the treatment plan. In the set-up procedure, the primary aim is setting up the equipment and patient according to the treatment plan. Often portal images, i.e. images based on the treatment beam itself, are used to verify the treatment and monitor its reproducibility. Furthermore, e.g. in vivo dosimetry or related techniques may be used to check the delivered radiation dose in the target volume and/or in adjacent tissues, particularly in organs at risk. If the measured data corresponds to the planed position in the treatment plan, the actual radiation therapy treatment may be safely initiated. As a consequence of the patient set-up, positioning and simulation procedure, the total treatment takes considerably longer time, generally at least 5-10 minutes and often more, than the actual irradiation. In addition, if some divergence between the measured and calculated data is detected during the set-up and simulation and the divergence exceeds the tolerance margin, the treatment set-up should be adjusted. This may in some cases simply be a correction of some set-up parameters but also larger adjustments requiring a renewed treatment planning process with new anatomical information from a renewed diagnostic procedure. When, a new treatment plan is needed, a renewed treatment simulation may also be needed, increasing the time to treatment by a day or two. Thus, the time, during which a radiation machine actually is employed for irradiating a patient, constitutes a small portion of the total time, during which the machine is occupied. This of course leads to poor utilization of the costly radiation machines and equipment and that fewer patients can be treated during a given period of time. This problem will be worsen further in cases where the patient undresses in the treatment room, the patient feels uncomfortable and wants to talk to the therapy assistants about various problems with the treatment, etc. A possible solution could be to perform the simulation procedure using a dedicated radiation simulation machine and not the actual radiation treatment machine. However, although the designs of the patient couches and the two machines used for the simulation and the treatment, respectively, are similar, it may be more difficult to correctly simulate the treatment using a different machine and possibly a different couch top. This is due to problems with positioning the patient exactly in the same way on two different couches, even though the couches may have the same design. In addition, tissue and organs, including the target volume with a tumor, are deformable elastic structures and their positions relative to reference points used in the treatment plan are not rigid, but may change depending on e.g. posture of the patient, filling degree of bladder, respiratory motion, etc. Therefore, although the reference points may be aligned correctly during the treatment relative those during the simulation, the target volume may be misaligned. In the patent specification U.S. Pat. No. 6,683,318, a therapy system adapted for cancer treatment using light ion radiation beams is disclosed. The therapy system includes an ion source providing light ions to an accelerator system including a synchrotron. An ion beam transport system guides an extracted high energy beam from the synchrotron into three different treatment rooms. In a first treatment room, a static gantry provides horizontal ion beam irradiation. In the remaining two treatment rooms, a respective rotatable isocentric gantry is arranged. Although, this therapy system is using a single ion source and beam accelerator system for the three gantries, the above-identified problems in terms of (low) patient throughput and cost-effectiveness are still present for the individual gantries of the therapy system. Furthermore, many treatments with charged particles are not made using the isocentric set-up principle, such as during electron, proton or light ion therapy, where generally a fixed SSD treatment with fixed distance between source and patient surface is performed, making isocentric treatment units less important. The present invention overcomes these and other drawbacks of the prior art arrangements. It is a general object of the present invention to provide a radiation system with an excentric gantry. It is another object of the invention to provide a radiation system with a gantry that can provide radiation beam delivery in multiple treatment rooms arranged around the gantry. It is a particular object of the invention to provide a radiation system that can direct a clinical radiation beam to a subject in a first treatment room while simultaneously a simulator and beam set-up part of the radiation system is used for irradiation set-up and simulation for other subjects in adjacent treatment rooms as a preparatory process to the irradiation. It is another particular object of the invention to provide a radiation system having integrated radiation treatment and simulator functionality that can be used in multiple treatment rooms. These and other objects are met by the invention as defined by the accompanying patent claims. Briefly, the present invention involves a radiation system or machine with an excentric gantry that can be used for irradiating subjects in multiple irradiation or treatment rooms. With such a gantry it is possible to irradiate, i.e. deliver radiation treatment doses, to a first subject in a first treatment room while simultaneously performing a treatment set-up and simulation procedure for at least a second subject in a second treatment room using the same radiation treatment gantry. When the irradiation of the first subject and the treatment set-up involving the second subject are completed, the gantry can be turned for delivering radiation treatment doses to the second subject simultaneously as a treatment follow-up involving the first subject or a new treatment set-up involving a third subject is performed in the first treatment room. As a consequence, the capacity of the radiation system of the present invention in terms of the total number of subjects to be treated during a given period of time is much larger compared to machines of the prior art. In addition, more time is available for patient care in each treatment room both before and after each treatment occasion. The radiation system of the invention comprises a gantry adapted for arrangement in connection with multiple treatment rooms separated by radiation-shielding or -isolating separating members, e.g. radiation-shielding partitions (walls) and/or ceilings/floors. A radiation head is mechanically supported by the gantry and is movable relative the gantry between at least a first position for directing a radiation beam into the first treatment room and a second position for directing the radiation beam into the second treatment room. The gantry of the radiation system is preferably arranged in the intersection of the partitions and/or the ceiling/floor separating the multiple treatment rooms. The radiation system typically has a spherical or cylindrical design, allowing a radiation head to rotate in a dedicated spacing in the partitions and/or ceiling/floor. As a consequence, the radiation head delivering the radiation doses can be turned between the different treatment rooms and therefore irradiate subjects positioned in these different such rooms. The gantry could include a static gantry part attached to the separating members. In such a case, a movable gantry part is movably (rotatably) supported by the static gantry part. The radiation head is then preferably attached to this movable gantry part. In addition, each treatment rooms preferably comprises or has access to a simulator head with e.g. a light optical and/or diagnostic X-ray system, being able to simulate the therapeutic beam from the radiation head. These simulator heads could be arranged and move concentrically on the gantry. Thus, in each room, the low cost radiation simulator could be used for patient set-up before the radiation head is turned into the treatment room for the real treatment. Alternatively, a few simulator heads could move between rooms and assist in setting up the patient prior to the treatment. Very many different room configurations can be anticipated from the basic configuration of the radiation system in the partitions and/or ceiling/floor of the rooms. Depending on where a treatment room is located around the central excentric gantry, typically 30-60° oblique lateral anterior, posterior and/or straight vertical and/or horizontal beam directions are possible. It is even possible to use two excentric gantries in such configuration so that multiple treatment portals can simultaneously be directed onto one and the same subject either as oblique lateral, parallel opposed or perpendicular beam combinations. The excentric gantry of the radiation system of the invention preferably comprises a beam scanning and bending system, such as a magnet-based system, adapted for scanning and bending an incident radiation beam onto a subject in form of a narrow pencil beam. This scanning and bending system, or at least a portion thereof, is then rotated as the radiation head and possibly gantry of the radiation system is turned between different rooms. A bending magnet of the bending system could be laminated to allow fast field changes between accelerator pulses but could also be super-conducting to minimize the bending radius of the magnet. Due to the design of the bending and scanning system, the excentric gantry is well adapted for usage with light ions from protons and upwards to carbon and oxygen ions, including for example protons, deuterons, tritium and helium, lithium, beryllium, boron, carbon and oxygen ions. The invention offers the following advantages: Can be efficiently used for irradiating several subjects in different treatment rooms; Enables treatment set-up, simulation and/or the subject exit process to be conducted in some rooms while the actual radiation treatment is simultaneously performed in another room; Increases the capacity in terms of the total number of subjects that can be irradiated during a given period of time; Enables usage of light ion radiation with a compact gantry design and small bending magnet radius and flexible beam direction selection; Reduces the installation cost substantially since one low cost device with adjustable beam directions can advantageously be used in multiple treatment rooms instead of multiple expensive isocentric devices with fixed beam line configurations; and Allows full attention to be paid to each patient with regard to individual care and accurate patient set-up without stress for the therapy assistance team. Other advantages offered by the present invention will be appreciated upon reading of the below description of the embodiments of the invention. Throughout the drawings, the same reference characters will be used for corresponding or similar elements. The present invention relates to a radiation system or machine with a so-called excentric gantry design that can be used for providing a radiation beam into multiple, i.e. at least two, irradiation or treatment rooms. The gantry is adapted for arrangement in connection with the multiple treatment rooms, which are separated by radiation-isolating or -shielding separating members. The gantry mechanically supports a radiation head. This radiation head is movable relative the gantry (and the separating members) between different positions for directing a radiation beam into the different treatment rooms. With such a gantry design it is possible to irradiate, e.g. to deliver radiation treatment doses to a first subject or patient, in a first treatment room while simultaneously allowing preparation for treatment (treatment set-up), simulation or performing treatment follow-up procedure for at least a second subject in another treatment room using the same radiation delivery system and gantry. As a consequence, the capacity of the radiation system of the invention, in terms of the total number of patients to be treated for a given period of time, is much larger compared to the prior art machines, e.g. the radiation machines of FIG. 1. In the following the invention will be described with reference to a radiation therapy system, delivering irradiation doses to a patient for the purpose of treatment, and then primarily curative radiation therapy, i.e. to eradicate a tumor. This radiation therapy system could also be employed for palliative radiation therapy, where the aim is generally to improve quality of life of the patient by maintaining local tumor control, relieve a symptom or prevent or delay an impending symptom, and not necessarily to eradicate the tumor. However, the radiation system could alternatively be employed for other radiation purposes, such a single dose radiation therapy, radiation diagnostics or radiation processing. In addition, the radiation system could be used for combined radiation therapy and diagnosis. In the latter case, the radiation head can deliver both a (high-energy) radiation treatment beam and a (low-energy) radiation diagnostic beam. Actually, the radiation system according to the present invention can be applied for any radiation purposes, where it is desired to direct a radiation beam onto an object or patient in a treatment room simultaneously as a radiation simulation, set-up or follow-up is performed in a (neighboring) treatment room involving another object or patient, to be subsequently irradiated or already has been irradiated, using the same radiation gantry. FIG. 2 schematically illustrates a radiation system or machine 1 of the invention with an excentric gantry 100 that is able to irradiate subjects or patients 50-1 to 50-4 in four different treatment rooms 61 to 64. Thus, the radiation gantry is arranged in connection with these four treatment rooms 61 to 64. In this embodiment, the gantry 100 is positioned in the intersection of the separating members, i.e. the walls or partitions 72, 74 and the ceiling/floor 71, 73, separating the four rooms 61 to 64. Note thus, that the treatment rooms 61 and 64 are positioned one floor beneath the treatment rooms 62 and 63. The separating members 71 to 74 separating the relevant treatment rooms 61 to 64 have radiation-shielding properties. Thus, the separating members 71 to 74 preferably prevent the treatment radiation beam 110 from leaking from the currently irradiated treatment room 61 to the other treatment rooms 62 to 64. As a consequence of these radiation-shielding separating members 71 to 74, (medical) personnel and patients can safely be present in a treatment room 62 to 64 even when the radiation system 1 irradiates 110 a patient 50-1 in an adjacent treatment room 61. In other word, the separating members 71 to 74 stop (absorb) the treatment radiation 110 so that the leaking radiation levels in the adjacent treatment rooms 62 to 64 are within defined safety margins. The choice of material and material thickness to use for the separating members 71 to 74 depend on properties of the employed treatment radiation 110, e.g. the energy level of the treatment beam 110, the type of radiation used, etc., and can be non-inventively determined by the person skilled in the art. Suitable materials for the separating members 71 to 74 include, but are not limited to, concrete, borated polyethylene and lead. Also the material surrounding and enclosing the radiation transport system, guiding the radiation into the radiation head 120, preferably has good radiation-shielding properties. The radiation gantry 100 typically has a spherical or cylindrical design, allowing a radiation head 120 to rotate in a dedicated spacing in the walls 72, 74 and ceiling/floor 71, 73. In the figure, the gantry 100 is directed to irradiate 110 a target volume 55-1 in a first patient 50-1 positioned on a treatment couch 40-1 in a first treatment room 61. The radiation system 1 preferably also comprises a radiation shielding 150, preferably a rotary radiation shielding, and bending magnet (see FIGS. 7-11) to deflect the radiation beam 110 into the currently used treatment room 61 and prevent it from reaching the other treatment rooms 62 to 64. In addition, the treatment rooms 61 to 64 preferably comprise or have access to simulator heads 200-2 to 220-4 with a light optical and/or X-ray system that is able to simulate the therapeutic beam 110. These simulator heads 200-2 to 200-4 could be arranged and move on a rail just outside of the shield 150. Thus, in the rooms 61 to 64, the low cost simulator 200-2 to 200-4 could be used for patient set-up before the radiation head 120 is turned into the treatment room for treatment operation. It is anticipated by the present invention that these simulator heads 200-2 to 200-4 instead, or alternatively, can be employed for treatment follow-up purposes. In the figure, three patients 50-2 to 50-4 are positioned on a respective treatment couch 40-2 to 40-4 and are currently subject to a treatment set-up (follow-up) and simulation procedure using the simulator heads 200-2 to 200-4 and simulator beams 210-2 to 210-4. In order to obtain maximum accuracy in the patient set-up (typically within 0.5-1 mm), a stereotactic treatment couch 40-1 to 40-4 is preferably used in the treatment rooms 61 to 64. Such a couch 40-1 to 40-4 is then automatically positioned and individually adjusted to each patient 50-1 to 50-4. Note that in the present invention one and the same couch 40-1 to 40-4 can be used both for the patient set-up (and follow-up) procedure and the treatment and diagnostic imaging activities. It could be possible that the treatment rooms 61 to 64 are equipped with one simulator head 200-2 to 200-4 each. Alternatively, two or more rooms 61 and 62 could share a single common simulator 200-2. In such a way, the simulator 200-2 is preferably able to move, e.g. by means of a rail system, in a dedicated gap 80 in the floor/ceiling 71 (or wall) separating the rooms 61 and 62. Once, the first patient 50-1 is treated and the set-up and simulation procedures are finished in another room, the gantry 100 may be turned so that the radiation head 120 now can deliver radiation doses to another patient 50-2 in another treatment rooms 62. This scenario is illustrated in FIG. 3, where the radiation head 120 (and possibly the gantry 100) has been rotated to irradiate a target volume 55-2 of a second patient 50-2. The simulator head 200-2 of the room 62, where the patient 50-2 currently is being irradiated using the treatment beam 110, is moved away either to one side of the room 62 in order to allow the radiation head 120 to irradiate the patient 50-2 or to the neighboring treatment room 61 for usage therein in treatment simulation. In the first room 61, a new patient 50-1 may be positioned on the couch 40-1 and a set-up procedure and simulator using a simulator head 200-2 can be performed. Alternatively, a treatment follow-up and subsequently patient exit procedure can be performed involving the previously irradiated patient. Generally, the sequence of events taking place in a given treatment room is as follows. Firstly, the equipment (couch and radiation and positioning equipment) in the room is readjusted in order to prepare for a next patient to be treated. A patient set-up is then conducted where the patient is accurately positioned on a couch, preferably a stereotactic couch, using different patient positioning systems, e.g. laser-based positioning systems, such as a patient positioning system described in the international patent application WO 2004/000120. Once the patient is accurately positioned on the (stereotactic) couch, a treatment simulation is then performed. During this simulation a light optical and/or diagnostic X-ray system in a simulator head of the radiation system of the invention is used. Thereafter the actual treatment can be performed. Since the treatment set-up normally can take at least 5 to 10 minutes and the actual treatment of the patient is much quicker, about 1-2.5 minutes, the treatment rooms 61 to 64 and patients 50-1 to 50-4 will generally have access to the therapeutic beam about every 10 minutes. For the gantry design of FIGS. 2 and 3 this implies that up to (and sometimes, especially for simple treatment, more than) 6×4=24 patients can be treated per hour in a very busy therapy center, and still allowing ample set-up time and patient care in each treatment room 61 to 64. This should be compared to the corresponding capacity of a prior art (isocentric) radiation machine, which typically maximally can treat up to 4 to 6 patients during that one hour time period. In addition, in this embodiment of the invention, a single radiation gantry with a single beam transport and scanning system can be employed for both treatment radiation and treatment set-up and simulation in multiple treatment rooms. Very many different room configurations can be anticipated from the basic four rooms configuration illustrated in FIGS. 2 and 3. Depending on which quadrant a treatment room is located around the central excentric gantry, typically 30-60° oblique lateral anterior (room 61 and 64) or posterior (room 62 and 63) beam directions are possible. With light ions it is very convenient to treat a patient requiring 2-4 beam portal directions in his treatment plan by one beam portal per day and, thus, sequentially use the different treatment rooms of FIGS. 2 and 3 as required by the beam directions. Thus, one and the same patient can, at different irradiation occasions, be irradiated in different treatment rooms, thereby receiving a treatment radiation beam from different incident angles and directions. It is also possible to have straight vertical and/or horizontal treatment beams in some rooms, possibly at the same time as other rooms use obliquely incident beams. FIG. 4 schematically illustrates this situation with four different treatment rooms 61 to 64 having access to one common radiation system 1 with excentric gantry design 100 according to the invention. Due to the position of the gantry 100 in the ceiling 71 relative a patient 50-1 on a couch 40-1 in room 61, straight vertical treatment beams 110 are possible. Similarly, for a patient 50-3 in the treatment room 63 posterior vertical beams are possible. However, in another treatment room 64 the beams will obliquely incident into a patient. In the figure, the patient support 40-4 of this room 64 is empty, schematically illustrating the principle with the patient entry/exit procedure. Correspondingly, if the radiation system 1 and the gantry 100 basically are arranged in a wall 72 between treatment rooms 62 and 63, a patient 50-2 may be vertically irradiated. It is anticipated by the invention that an excentric gantry and radiation system of the present invention could be adapted for only radiating horizontally, only vertically, only obliquely or a combination of horizontally, vertically and/or obliquely. In this embodiment of the invention, the simulator heads 200-1 to 200-3 have been arranged in dedicated hollows or portions of the gantry 100 or in the radiation shield 150 of the gantry 100. These hollows may be fixed but is preferably movable or rotatable so that the simulator heads 200-1 to 200-3 can be moved away (possible between the rooms 61 to 64) to allow space for the radiation head 120 to irradiate a patient. It is also possible to combine vertical, anterior and posterior beams in an excentric gantry with six surrounding treatment rooms by combining the solutions from FIGS. 2 or 3 and 4. Such a gantry design results in both oblique lateral anterior and posterior beams in four treatment rooms as well as parallel opposed vertical beams in two rooms. FIG. 5 is an illustration of a radiation system 1 and excentric gantry 100 according to the present invention that are able to provide (treatment) radiation 110 in multiple radiation treatment rooms 61 to 68 positioned in up to three different floors. Thus, three treatment room 61, 67, 68 are in a first floor, a second floor includes two treatment rooms 62, 66 with the radiation gantry 100 basically positioned therebetween. A third floor then includes the remaining three treatment rooms 63 to 65. In this embodiment, the gantry 100 is arranged in connection with two radiation-shielding wall-pairs 73, 78 and 74, 77 and two radiation-shielding ceiling-floor pairs 71, 76 and 72, 75 separating the eight treatment rooms 61 to 68. As a consequence, up to seven patients 50-2 to 50-8 can be involved in treatment set-up, simulation or follow-up procedures, possibly employing radiation simulation or diagnosing heads 210-2 to 210-8, simultaneously as a patient 50-1 is irradiated using the treatment radiation beam 110 of the radiation system. It is also possible to rotate a patient 180° in the horizontal plane in a given treatment room, such as in room 61 in FIG. 5, to obtain during one treatment session a pair of anterior oblique and lateral beam portals on the tumor. This is one of the most efficient treatment configurations for shallow to half deep tumors. For deep therapy, parallel opposed anterior-posterior beams may be most efficient, particularly with higher LET (Linear Energy Transfer) ions like carbon and oxygen, as could be effectively delivered in rooms 64 and 68 of FIG. 5. With reference to FIG. 6, it is even possible to use a radiation system 1 with multiple, i.e. at least two, excentric gantries 100-1 and 100-2 in such configuration that multiple treatment portals can simultaneously be directed onto one and the same patient 50-4 either as oblique lateral (as in the figure), parallel opposed and/or perpendicular beam combinations 110-1 and 110-2. Thus, in the configuration of FIG. 6, patients 50-3 and 50-4 in room 63 and 64 can be irradiated with beams 110-1 and 110-2 from the radiation heads 120-1 and 120-2 of the two excentric gantries 100-1 and 100-2. In the figure, the remaining four rooms 61, 62, 65 and 66 only have access to one of the gantries 100-1 or 100-2. It is anticipated by the invention that more than two excentric gantries according to the present invention may be arranged in a common configuration so that at least two or more gantries are able to irradiate a patient in one of the treatment rooms. It is also possible, by changing the arrangement of the gantries in the walls and ceiling/floor, to combine the gantry arrangement of FIG. 4 with the arrangement in FIG. 6. The incident radiation to these at least two gantries of the radiation system of the present invention can originate from different radiation sources. Alternatively, a common radiation source, possibly including an ion source, accelerator system (with a synchrotron or cyclotron), beam guiding and splitting system, can be used for the at least two gantries. An example of a suitable ion source, accelerator and beam guiding system that can be used according to the present invention is disclosed in the patent specification U.S. Pat. No. 6,683,318. The beam splitting system can be realized by a septum, e.g. in the form of a thin conducting foil, arranged in the path of the beam. By applying a (high) current to the septum, the resulting induced magnetic field can be used to divide the incoming ion beam into at least two outgoing ion beams. Each such outgoing beam can be brought, via a beam guiding system, to a respective excentric gantry of the invention, e.g. as illustrated in the configuration of FIG. 6. The radiation system of the present invention could be designed so that the radiation gantry is static and the radiation head is movable (rotatably) attached to this static gantry to be able to move (rotate) between the multiple treatment rooms. In an alternative embodiment of the present invention, the gantry could include a static gantry part or portion and a movable gantry part or portion. The radiation head is then mechanically supported by the movable gantry part. In such a case, the static gantry part could be attached to or in connection with the separating members of the treatment rooms. In such a design, the movable gantry part and radiation head are movable (rotable) relative the static gantry part (and the separating members) between the treatment rooms. In FIGS. 2-6, the radiation system according to the present invention has been disclosed with reference to arranging the radiation gantry in connection with a separating member separating at least some treatment rooms positioned at different floors (levels). However the present invention is not limited thereto. For example, the gantry of the radiation system can be arranged in the (radiation-shielding) partition separating two treatment rooms positioned on the same floor. In such a case, the radiation system can direct a radiation beam into at most two different treatment rooms. However, a stand-alone gantry with a general cylindrical or column- or pillar-like design positioned in a large room or hall is also possible. In such a case, this large room can be screened or divided into multiple irradiation stations or (smaller) rooms, separated by radiation-shielding separating members (partitions). Thus in this embodiment, a respective first short end of the separating members is connected to or at least close to the lateral surface of the cylindrical gantry. The separating members then protrude (possibly radially) from the gantry to define the different radiation-shielded treatment rooms. The radiation head can be rotatably attached to the gantry or the radiation head and the movable gantry part is rotatably attached to the static gantry part, in turn arranged on the floor. Depending on the height of the patient couches in the treatment rooms and the output angle of the radiation beam, oblique or horizontal irradiation is then possible. FIG. 7 illustrates a perpendicular cross section view of an embodiment of a radiation system 1 with an excentric gantry design 100 of the invention. The incident radiation beam may come from a nearby located radiation source, such as a radiation source arranged in a room next to the gantry 100 and using bending magnets to direct the radiation into the gantry. It is also possible to use a radiation source that is arranged directly on the gantry 100, or a relatively far-placed radiation source, e.g. synchrotron, (see e.g. U.S. Pat. No. 6,683,318) or cyclotron, that can deliver the required radiation to several different treatment units and excentric gantries 100. Herebelow the invention will be described with reference to a radiation therapy system comprising a (pencil) beam scanning system 104-106, 122 for irradiating a patient 50. Thus, magnetic fields provided by the scanning system 104-106, 122 are used to control the charged particles in the radiation beam. In this beam controlling, the spot-size of the beam can be adjusted and scanned over a treatment area in the patient 50. By a variation of the scanning speed and the beam intensity any desired dose distribution within the target volume 55 can be generated with a minimum extra dose delivered to healthy tissue. However, the invention is not limited to such pencil-like radiation therapy systems and scanning techniques. According to a preferred embodiment of the invention, the incident radiation beam from the (remote) radiation source (not illustrated) first enters quadrupoles 102 for focusing the beam. Thereafter, the beam preferably enters a scanning magnet 104. This magnet 104 deflects the beam and gives it a scanning motion in the plane of the drawing. The beam emerges from the scanning magnet 104 as if it came from an effective scanning center typically near the middle of the magnet 104. The beam scanning in the plane of the drawing is then bent or deflected in a bending or deflecting magnet 106 for directing the incident beam down to the radiation head 120 and subsequently into the target volume 55 of a patient 40. The bending magnet 106 could be laminated to allow fast field changes between accelerator pulses but also super-conducting to minimize the bending radius. Due to the beam bending functionality of the magnet 106, the beam enters the radiation head 120 and a second scanning magnet 122. This magnet 122 has the ability to scan or deflect the beam in a plane transversal to the plane of the drawing, i.e. in and out of the plane of the drawing. The beam could then enter a collimator 124 that is arranged to prevent radiation outside the intended scanning beam to continue down to the patient 50. An optional transmission monitor 125 could be provided below the collimator 124 for registering the amount radiation passing from the collimator 124. The scanning system of FIG. 7 could also be switched off to use a regular dual or single scattering foil system in order to get simple uniform beams. Before the radiation beam 110 leaves the radiation head 120 it preferably passes a second collimator 126. This collimator 126 is preferably of a multi-leaf collimator type. Such a multi-leaf collimator type comprises a plurality of pairs of opposed elongated, curved or flat, in cross-section wedge shaped leaves, each adjacent leave arranged side by side and such that a fan-shaped configuration which converges towards an apex of the effective radiation source 127. Contrary to the present invention with preferably a single source 127 conventional scanning systems with two consecutive dipole scanning magnets will have different effective source locations in the two scanning planes. The leaves of the collimator 126 are mounted for (combined) rotational and/or translational movement. This dynamic multi-leaf collimator 126 can be used to protect normal tissues lateral to the tumor, i.e. the target volume 55, at the same time as the magnetic field of the bending magnet 106 is rapidly adjusted to the energy required at each scan position. Normally, the energy remains fixed during the scanning of the beam 110 at a certain depth in the patient 40. The scanned beam 110 typically covers a 30 cm×30 cm field size on the patient 40. If a transmission monitor 125 is provided in the radiation head 120, this monitor could continuously follow and interlock the motion of the scanned beam. The radiation machine 1 of the invention with an excentric gantry design as illustrated in FIG. 7 is well adapted for usage of a radiation beam of light ions, i.e. from protons and upwards, e.g. helium, carbon or oxygen ions. Such ions are very effective in treating patients with cancer disease. Since they have favorable physical and biological properties that can be exploited for developing improved treatment techniques in comparison to conventional proton beams, light ion beams offer a unique combination of several advantages including a high physical selectivity and higher biological effectiveness in the Bragg peak. As is well known in the art, light ions require very large bending radii (up to several meters). Prior art radiation gantries and machines providing light ion dose delivery suffer from high installation cost. In addition, such prior art machines have several large sized bending magnets and their pool gaps, which are required to scan the beam in both planes before the bend and obtain a gantry adapted for delivering ion beams in arbitrary orientations in the treatment room. However, the bending magnet 106 used in the present invention can have a small gap and smaller radius and consequently smaller overall size than the prior art used magnets. This results in a compact pencil beam scanning system comprising the scanning and bending magnets 104, 106, 122 that can provide (30 cm×30 cm) beams to several treatments rooms placed around the excentric gantry 100. Although the gantry design is suitable for usage with light ion radiation it can also be used for any charged particle or even neutral particles like neutrons and photons by first scanning the primary deflected proton or deuteron and electron beams to generate scanned neutral beams, see e.g. the patent specification U.S. Pat. No. 4,442,352. In FIG. 7, the gantry 100 has a static gantry part 140 attached to the floor/ceiling 71 separating the two adjacent treatment rooms 61, 62. An inner movable gantry part 130 with attached or integrated radiation 120 is movably, here rotatably, supported by the static part 140. This movable support can be realized using conventional gear solutions and bearings. In order to increase the stability of the gantry 100, the radiation head 120 can be supported by the movable gantry part 130 the right-hand side (as illustrated in the figure) and to the left-hand side. In such a case, the gantry 100 typically would include two static parts 140, each arranged in the floor/ceiling 71 but on either side of the rotating radiation head 120. When the radiation machine is to be used for treating a patient in another treatment room 62, the gantry 100 (the movable gantry part 130) is simply turned or rotated, thus resulting in a rotation of the scanning and bending system 102-106, 122 and the radiation head 120. FIG. 8 illustrates this principle, where the excentric gantry 100 of the radiation system 1 of FIG. 7 is turned from irradiating a patient in a first treatment room 61 to irradiating a second patient 50 lying on a treatment couch 40 arranged in a second treatment room 62. As is seen in the figure, due to the rotation of the gantry 100, the bending magnet 106 now directs the incident beam towards this second treatment room 62. In this way a single scanning, collimation, beam bending and angular adjusting system can be used in several treatment rooms, which significantly reduces the cost for the whole installation. FIG. 9 is a corresponding cross-sectional illustration of another embodiment of a radiation system according to the present invention. In this embodiment, the gantry 100 is mainly arranged in a radiation-shielding wall or partition 71 separating two neighboring treatment rooms 61 and 62. This gantry embodiment 100 is in particular configured for providing horizontal beams 110 to the treatment volume 55 of the patient 50. The including units 124-126 of the radiation head 120 and the beam processing (guiding and scanning) system 104-106, 122 are similar to the corresponding units discussed above in connection with FIG. 7 and are not further described. Although the rotation axes of the excentric gantries illustrated in FIGS. 2 to 8 are horizontal or vertical as illustrated in FIG. 9, the present invention is not limited thereto. The rotation axis can have any angle from vertical to horizontal, somewhat dependent on how the beam is extracted from the accelerator and the range of variability needed clinically. FIG. 10 illustrates a cross-section of a portion of radiation system 1 with a gantry 100 providing oblique beams 110. Compared to the radiation system of FIG. 7, in the present embodiment the treatment beam 110 is deflected by the bending magnet 106 less than 900 in order to save power and to get oblique beams in multiple treatment rooms 61 and 62. Angles as low as 30° up to some 60° could be useful in special cases. The including units of the gantry 100 and radiation system 1 correspond to the units discussed above in connection with FIG. 7. FIG. 11 illustrates another possible design of the internal units of the beam scanning and bending system 104-106, 122 of the excentric gantry 100 of the present invention. This embodiment minimizes the diameter of the rotary gantry part 130 including the magnet 106 and treatment head 120. Similarly to the embodiment illustrated in FIGS. 7 to 10, the incident (light ion) radiation beam, preferably, first enters quadrupoles 102. The beam then enters a bending and scanning magnet 104 that scans the beam in the plane of the drawing. The scanned beam is then bent in a (super-conducting) bending magnet 106. The bending magnet 106 preferably directs the beam into a second scanning magnet 122. This scanning magnet 122 and the collimators 124 and 126, where discussed in connection with FIG. 7. In order to really minimize the gantry size and maximize the cost effectiveness, the radiation shielding 150 may also be included, calling for a somewhat smaller deflection in the magnet 104 than shown in the FIG. 11. It will be understood by a person skilled in the art that various modifications and changes may be made to the present invention without departure from the scope thereof.
	claims	1. A single-leaf X-ray collimator comprising:at least one planar collimating leaf member disposed along a path of X-rays, the planar collimating leaf member comprising at least one non-circular continuously elliptical collimating aperture therewithin,wherein the planar collimating leaf member is configured to rotate about at least one of a horizontal or vertical direction, anda driving means for tilting the at least one planar collimating leaf member relative to the path of X-rays in the least one of a horizontal or vertical direction. 2. The single-leaf collimator according to claim 1 wherein the collimating aperture provides improved collimating efficiency. 3. The single-leaf collimator according to claim 1 further comprises at least one auxiliary leaf member disposed along the path of X-rays. 4. The single-leaf collimator according to claim 3 wherein the auxiliary leaf member is provided in combination with the collimating leaf member. 5. The single-leaf collimator according to claim 3 wherein the collimating leaf member comprises a source side and an imager side for X-rays, wherein the auxiliary leaf member is disposed at the source side of the collimating leaf member. 6. The single-leaf collimator according to claim 3 wherein the auxiliary leaf member comprises a size that is predetermined to cover the entire field of X-rays at a distance from the source side. 7. The single-leaf collimator according to claim 3 wherein the collimating leaf member and the auxiliary leaf member are constructed of an X-ray attenuating material. 8. The single-leaf collimator according to claim 1 wherein the driving means further comprises a DC servomotor. 9. The single-leaf collimator according to claim 8 wherein the auxiliary leaf member further comprises the driving means. 10. A single-leaf collimator comprising:a housing;a collimating member operable to collimate a beam of X-rays within said housing; anda driving means operably coupled to the collimating member,wherein when the collimating member is rotated in at least one of a horizontal or vertical direction by the driving means, the X-ray beam is collimated to a non-circular continuously-elliptical shape. 11. The single-leaf collimator according to claim 10 further comprising an auxiliary leaf member in combination with the collimating member. 12. The single-leaf collimator according to claim 11 wherein the auxiliary leaf member is adapted for sliding along the path of X-rays, in combination with the collimating member. 13. The single-leaf collimator according to claim 11 wherein the auxiliary leaf member is configured to cover about entire field of X-rays. 14. The single-leaf collimator according to claim 11 wherein the auxiliary leaf member is configured to allow passage of X-rays therethrough to the collimating member. 15. The single-leaf collimator according to claim 14 wherein the collimating member and the auxiliary member are constructed of an X-ray attenuating material. 16. The single-leaf collimator according to claim 11 wherein the auxiliary leaf member further comprises the driving means. 17. The single-leaf collimator according to claim 11 wherein the auxiliary leaf member further comprises:an auxiliary leaf member for collimating, the auxiliary leaf member being substantially large to cover an entire field of the X-ray beam, in a rotated position and allow passage of the X-ray beam only through an aperture of the collimating member. 18. The single-leaf collimator according to claim 10 wherein the collimating member comprises a source side and an imager side for X-rays, wherein the auxiliary leaf member is disposed at the source side of the collimating member. 19. The single-leaf collimator according to claim 10 wherein the driving means further comprises:a hydraulic actuator. 20. The single-leaf collimator according to claim 10 wherein the driving means further comprises:a DC Servo motor.
056051719	summary	BACKGROUND OF THE INVENTION Tritium is a beta emitter with a low energy emission spectrum. The maximum electron energy is 18.6 keV with a mean around 5.7 keV. Tritium has been employed in the past in gaseous form as the leading isotope for radioluminescent applications, such as emergency signs in aircraft and hospitals where maintenance-free/absolute reliability needs exist. These lighting devices operate by having tritium in gaseous form next to a phosphor material. The beta emission from the tritium causes optical excitation of the phosphor (such as zinc sulphide) which provides the light emission. Limitations of these technologies include the fact that the only useful beta emitters are those that can be located near a phosphor. Most of the emitted electrons will fail to reach the luminescent materials, and without proper configuring, scale-up of the intensity/power level will be prohibitive. This is a direct result of the short emitted electron range in both gaseous and condensed phase media from a low energy beta emitter, such as a tritium. The range can vary from 6 mm in air to 6 microns in water to lengths that are substantially less than these distances in solid materials. But safety and environmental reasons dictate that only the lowest energy emitters (of which tritium is a prime candidate) be considered. Another limitation is safety/environmentally related. Currently most utilization of tritium is in the gaseous form. In case of containment failure, the escape of tritiated gas into the environment is rapid and difficult to contain. SUMMARY OF THE INVENTION Therefore, an object of the present invention is to circumvent both limitations via the covalent bonding of tritium onto the interior surfaces of porous silicon. In effect the beta emitters are located at most a few angstroms from the luminescent center, namely the silicon species. Instead of a near-surface effect, the beta emitters have now become volumetric sources that are dispersed microscopically throughout the the entire porous silicon material. Practically all the beta emitters and all the silicon materials can interact with each other to the extent that the issue of the short range of the beta particles from tritium becomes irrelevant. Intensity/power level scale-up is now proportional to the volume of the device rather than to the area as in current technologies. The fact that the tritium is now encapsulated in a solid state environment with the tritium itself in condensed matter form (as interior surface chemisorbed species) renders its escape into the environment several orders of magnitude less likely than the gas-based configurations used in current technology. Thus, the present invention of a tritiated porous silicon (TPS) material provides an optical power source whose output will scale with the volume of the device, i.e., the best scaling possible, and in an embodiment where the tritium is much better contained. In this context another object of the invention is to use TPS for a self-powered solid state light source with unique characteristics for lighting applications, such as TPS as a stand-alone optical/power source in silicon-based optoelectronic device technologies or in combination as photovoltaic devices. The invention consists of certain novel features and a combination of parts hereinafter fully described, illustrated in the accompanying drawings, and particularly pointed out in the appended claims, it being understood that various changes in the details may be made without departing from the spirit, or sacrificing any of the advantages of the present invention.
048308161	summary	BACKGROUND OF THE INVENTION The invention relates to a getter trap for removing hydrogen and oxygen from liquid metal, such as sodium, and, more particularly, to such a getter trap for removing hydrogen and oxygen impurities from the liquid metal coolant of liquid metal nuclear reactors. Liquid metal nuclear reactor systems typically include primary and secondary liquid metal coolant loops. The primary liquid metal coolant, such as sodium, is heated as it passes through the nuclear core of the liquid metal nuclear reactor, and is subsequently cooled by indirect heat exchange with the secondary liquid metal coolant in a heat exchanger. Liquid metal circulating in the secondary coolant loop then passes through a second heat exchanger wherein the thermal energy from the secondary liquid metal coolant is used to heat water flowing in a third loop to produce steam, which, in turn, is used to drive an electricity generating device, such as an electrical generator. The secondary liquid metal coolant often contains undesired hydrogen and oxygen. The primary system also contains oxygen and hydrogen from the types of sources such as oxygen-bearing impurity gases in the argon or other cover gas, and the "dirt burden" on the surface of materials of construction. Hydrogen, such as that produced by corrosion in the third loop, passes into the secondary liquid metal coolant through the second heat exchanger. A typical hydrogen load in a secondary liquid sodium coolant loop in a 300 Megawatt power liquid metal nuclear reactor is 5432 kilograms of hydrogen per 30 years. Oxygen impurities are produced at a lesser rate than hydrogen impurities are produced. Thus, the secondary coolant loop often includes a means to eliminate the undesired hydrogen and oxygen from the secondary liquid metal coolant. One method of removing the hydrogen and oxygen impurities from liquid metal is the use of a cold trap. Cold trapping action depends upon the decreasing solubility of impurities with decreasing temperature. In a typical cold trap, a liquid metal system is cooled as it passes through a subsidiary system which includes a vessel where the impurities are precipitated and held as solid phases. In a conventional cold trap, the liquid metal enters the top of the trap from a heat exchanger and flows downwardly through the outer annulus, or "downcomer". The flow direction reverses at the bottom of a cylindrical volume packed with mesh. The liquid metal then flows upwardly through the cylinder and out through the end of the cylinder to an exit pipe. In general, nucleation of precipitate, and subsequent growth thereof, occur largely on the coldest surfaces of the cold trap. Typically, the coldest sections of cold traps are the bottom of the cold trap, the outer vessel wall, and the extended surface of the inlet section of mesh. The nucleation of precipitate and subsequent growth of the nuclei throughout these areas of the cold trap eventually restrict the flow path of the liquid metal through the cold trap and the cold trap is plugged even though only a small portion of the cold trap is filled with impurities. I disclosed a cold trap in my U.S. Pat. No. 4,291,865, which issued Sept. 29, 1981 to the assignee of the subject invention having a radial design for removing impurities, such as oxygen, from a liquid metal. U.S. Pat. No. 3,853,700 discloses a trap for carbon in liquid sodium that utilizes binary alloys of iron and from 0.5 to 30 weight percent of titanium, from 0.5 to 25 percent vanadium, or from 0.5 to 5 percent manganese. U.S. Pat. No. 3,993,453 relates to a getter trap of a composite with a substrate of a metal having a large coefficient of thermal expansion, such as nickel or nickel alloy, having a coating, which fractures upon heating, of zirconium or zirconium alloy for a nuclear fuel element. Getter traps have also been used to absorb hydrogen from liquid alkali metals. U.S Pat. No. 3,622,303 relates to such a getter comprised of a barrier of a first layer of iron, nickel, tantalum, columbium and their alloys, and a second layer of palladium, platinum and their alloys. U.S. Pat. No. 4,312,669 describes a gettering alloy for gases which include oxygen and hydrogen that is selected from the group consisting of an alloy of zirconium, vanadium and iron, whose composition, in weight percent, when plotted on a ternary diagram, 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; PA1 (c) 45 percent zirconium, 50 percent vanadium and 5 percent iron. PA1 (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. The disclosure of U.S. Patent No. 4,312,669 is incorporated by reference herein. SUMMARY OF THE INVENTION The present invention provides a getter trap to remove hydrogen and oxygen from a liquid metal, such as sodium. The getter trap includes an elongated, closed housing having an inlet at one end thereof and an outlet at the other end. A getter material is randomly disposed within the housing comprising a zirconium-containing substrate of hollow, tubular sections having a coating thereon of a gettering alloy of zirconium, vanadium and iron, whose composition, in weight percent, when plotted on a ternary diagram, lies within a triangle having as its corners the points defined by: Thus, as the liquid metal flows through the inlet into the housing and through the getter material, and is discharged from the housing through the outlet, hydrogen and oxygen impurities are removed from the liquid metal. The getter trap can include a second getter material comprising pellets of the gettering alloy. The invention provides a getter trap useful for the absorption of both hydrogen and oxygen that is effective at a temperature of 310.degree. C. or higher, which is typical of the temperature of liquid metal coolant passing through the secondary loop of a liquid metal nuclear reactor system. Thus, additional heat exchange to further cool and heat the liquid metal is not required, since it is not required to maintain a precise temperature for successful operation. The performance of the getter trap of the invention is not significantly affected by changes in temperature or flow performance. The invention further provides a getter trap utilizing a getter material which allows for a hydrogen loading of the zirconium-containing substrate to the theoretical maximum corresponding to the chemical formula ZrH.sub.1.3. Typical cold traps allow for hydrogen loading of less than half the theoretical maximum. When the auxiliary equipment associated with a cold trap and the different hydrogen density in the solid hydrogen-bearing phases in cold trap and getter trap are taken into account, the volume of the getter trap is about half that for a typical cold trap for the same hydrogen absorption capacity. The getter trap of the invention is particularly useful in a liquid metal cooled nuclear reactor system having a primary coolant loop, a secondary coolant loop, and a third coolant loop. The primary coolant loop includes a primary liquid metal coolant, a liquid metal nuclear reactor and a first heat exchanger. The primary liquid metal coolant flows through the liquid metal nuclear reactor, the first heat exchanger, and back to the liquid metal nuclear reactor. The secondary coolant loop includes a secondary liquid metal coolant, the first heat exchanger and a second heat exchanger, wherein heat from the primary liquid metal coolant of the primary coolant loop is transferred to the secondary liquid metal coolant through the first heat exchanger. The secondary liquid metal coolant flows through the first heat exchanger, the second heat exchanger, and back to the first heat exchanger. The third coolant loop includes a water coolant, the second heat exchanger and a steam driven device, wherein heat from the secondary liquid metal coolant of the secondary liquid metal coolant loop is transferred to the water coolant through the second heat exchanger. The water coolant flows through the second heat exchanger, the steam driven device and back to the second heat exchanger. A getter trap of the invention is disposed in at least one of the primary and secondary coolant loops.
053655567	abstract	A rack structure for storing channeled new fuel that awaiting a refueling operation in a nuclear reactor. The racks are designed to be floor mounted in a shallow fuel storage pool. Each rack comprises a pair of rows of fuel storage cells arranged back to back. The new fuel is loaded into the rack using a hoist, one fuel bundle being placed in each fuel storage cell. Each fuel storage cell has a latch which is mechanically linked to a contact plate at the bottom of the fuel storage cell. Each fuel bundle is latched in the fuel storage cell when the weight of the fuel bundle deflects the contact plate downward. Each fuel storage cell comprises an inclined channel which stores the fuel bundle assembly in an inclined position. The latch opens under the force of gravity in response to the fuel bundle assembly being lifted.
043404430	description	Referring to the drawings, auriferous rock 1 is fed to a hopper 2 which supplies it to a rock crusher 3 in which it is crushed into lumps 4 corresponding to a mesh size of some 5 cm. The stream of crushed rock leaving the crusher 3 is divided into a number of streams 5 only one of which is shown, which pass through a neutron irradiation assembly 6, to be described more fully later. Having been irradiated by neutrons generated within the assembly 6, each of the streams of lumps 4 of rock is caused to pass a .gamma.-ray detector assembly 7 which is arranged to detect any .gamma.-rays having an energy of 279 keV arising from the nuclear reaction .sup.197 Au (nn') .sup.197.sbsp.m Au, occurring in any gold contained in the lumps 4 of rock. Each lump 4 of rock is interrogated individually to establish whether its gold content lies above or below some predetermined concentration. For example, the critical concentration might be 5 ppm. In general it might be in the range 1 to 10 ppm. Downstream of the .gamma.-ray detector assembly 7 is a sorter 8 of a type which is well known in the art of material sorting, and which will not be described further. The sorter 8 is arranged to respond to signals from the .gamma.-ray detector assembly 7 to accept or reject for further processing each lump 4 of rock passing through it. Referring to FIG. 2, the neutron irradiation assembly 6 consists of a cylindrical body 21 made of lead which is surrounded by a biological radiation shield 22 which is made to be impervious to neutrons and to .gamma.-rays. In the body 21 there is a central bore 23 around the periphery of which there are positioned six tubes 24 made of boron. The tubes 24 extend throughout the length of the body 21. Each of the tubes 24 has a bore which is such that only a single stream of lumps 4 of rock can pass through the relevant tube. In the central region of the bore 23 in the body 21 there is a target 25 made of a material which will produce neutrons in response to bombardment by a beam of deuterons from a source which is not shown. Suitably, the target 25 can be made of a material which contains deuterons, or beryllium. The important thing is that the neutron source should be made of a material which does not produce neutrons which are energetic enough to excite fast neutron reactions in the constituent elements of the rock in which the gold is contained, i.e. aluminum, silicon, calcium, iron and oxygen. The target 25 will emit neutrons over a solid angle of 4II, but as the lumps 4 of rock pass through the maximum neutron field at 90.degree. to the direction of the neutrons, it can be arranged that while the neutron source energy equates to the maximum energy of the reaction EQU .sup.197 Au(nn').sup.197.sbsp.m Au, the neutron energy is below the threshold energies of (n,p) reactions in such as those previously mentioned which are likely to be present in the rock in high concentrations. In particular care should be taken to ensure that the neutron energy is below the threshold of the fluorine reaction EQU .sup.19 F(n,.alpha.).sup.16.sbsp.N.sbsb..beta. .fwdarw..sup.16 O(T.sub.1/2 =7.3 secs) This reaction generates .gamma.-rays having energies of 6.1 and 7.2 Mev. Although these are considerably greater than the 0.279 Mev from the .sup.197.sbsp.m Au, the half-life is the same, and the reaction could be the source of low-energy collided .gamma.-rays which would have the same decay pattern as those to be detected, and so interfere with the estimation of the gold content of the lumps 4 of rock, particularly if the fluorine is present in concentrations which are relatively high when compared to that of the gold. The .gamma.-ray detector assembly 7, which is not illustrated in detail, has six linear arrays of .gamma.-ray detectors, one for each stream of lumps of rock. The signals which operate the sorter 8, which again has six input channels, are derived from the combined output signals from each of the individual .gamma.-ray detectors appropriate to each channel. A neutron output of some 10.sup.11 n/s from the neutron source enables a lump of rock having a gold concentration of 1 ppm to be differentiated from one having a gold concentration of 2 ppm in one of three measurements, and a lump having a gold concentration of 2 ppm to be differentiated from one having a gold concentration of 5 ppm in 99 in 100 measurements.
063226936	description	DETAILED DESCRIPTION OF THE INVENTION Referring now to the drawings wherein like reference numbers designate identical or corresponding parts throughout the several views, and particularly, the FIG. 1 wherein the waste processing system in accordance with the present invention, generally designated 1, comprises a first mixing tank system 9 having at least one mixing tank 10. Each of the mixing tanks 10 have an agitation system situated within the tanks. The agitation system of the present invention both mixes and shears. The agitation system of the present invention includes shearing devices and shearing systems. The agitation system comprises at least one motor, a gear box and a plurality of blades. The motor can be a Blue Chip Motor manufactured by Marathon. The first separation tank may also have a grading situated across the top of the tanks. The size of the mixing tank 10 is dependent upon how fast the operator wants to process the waste and the size and type of particles of waste being processed. At least one modified centrifugal pump 5 is attached to the first mixing tank system 9. The modified centrifugal pump is modified by increasing the spacing between the holder and impeller of a centrifugal pump. The centrifugal pump may be a Magnum 250 Pump which is manufactured by Harrisburg, Inc. There are jet lines and re-circulation lines attached to the first mixing tank system 9. A first separation system 11 comprising at least one separation device 12 is attached to the first mixing tank system 9. The separation device 12 can be any vibrating screen such as shale shaker. The separation device may be a Scalping Shaker manufactured by Fluid Systems Corp. A manifold can be attached to the first separation device to control the speed and flow of the waste processing system 1. A slurry tank system 13 is connected to the first separation system 11. The slurry is tank system 13 comprises at least one slurry tank 14 having a shearing system situated within each tank 14. The shearing system comprises a gear box, at least one motor, a plurality of blades, a shearing mixer and gun lines. The motor may also be a Blue Chip Motor manufactured by Marathon. The shearing Mixer may be a Rotostat mixer manufactured by Admix. The gun lines may be Mud Guns manufactured by Harrisburg. At least one shearing pump 25 is attached to the slurry tank system 13. The slurry tank system 13 has a plurality of outlets whereby re-circulation lines and conduits can be attached. A manifold system is also attached to the slurry tank system 13. A second separation system 15 comprising at least one separation device 16 is attached to the slurry tank system 13. The separation device 16 in the second separation system is similar to the separation device in the first separation system 11. A second mixing tank system 17 comprising at least one mixing tank 18 is attached to the second separation system 15. The size of the mixing tank 18 is contingent upon the size and type of particles of waste being processed. The second mixing tank system 17 has an agitation system situated with the mixing tanks 18. The agitation system is similar to the agitation system of the first mixing tank system 9. A centrifugal pump 35 is attached to the second mixing tank system 17 and a manifold system is attached to the centrifugal pump 35. A holding tank 20 is attached to the second mixing tank system 17. The first mixing system 9, the first separation system 11, the slurry tank system 13, the second separation system 15, the second tank system 17 and the holding tank 20 are connected by conduits 7. The re-circulation lines and conduits 7 can be connected to any of the tanks using the manifold system. The holding tank 20 may have a plurality of compartments 20a-20d respectively and each compartment may have gauges to measure the amount of solids that are injected into the well. The holding tank 20 can also have a manifold system. At least one of the compartments can retain water to control the pressure within the compartments 20a-d. The waste processing system 1 can have a motor control room or a (SCR) service control room wherein control boxes, power sources, etc. are stored. Referring now to FIG. 2, the waste material is fed to the first 250 bbl mix tank 10 by a conveyor belt or by a pump 5. The main objective of this tank 10 is to fragment the clumps of clay pit solids and mix the solids with water. The tank 10 has two agitators within the tank and at least two centrifugal pumps 5. As the agitators rotated, the clumps of clay solids were battered by the paddles and broken up into smaller solids. The centrifugal pumps 5 are multifunctional. First, the pump sucks the solids from a bottom manifold and the solids are then pumped to the first stage separation device 12. To relieve the amount of flow, a circulating or manifold system was installed outside the tank 10. The circulation or manifold system functions to mix the solids in the 250 bbl tank 10. Second, the pump 5 is used as a back up pump and to assist in the mixing process within the tank. Third, the pump 5 may also function as a circulation pump. After the material is pumped over the first stage separation device 12, the device 12 equipped with large mesh screens functions to separate any large debris and unwanted metals out of the slurry mix before allowing the processable waste to enter into the slurry tanks 14. The debris is collected into an area to be disposed of in another manner or can be further processed. The processable waste is then drained into the slurry tanks 14 at a controlled rate. The slurry tank system 13 has valves to control the amount of flow or the direction of processed waste. Each slurry tank 14 is equipped with a centrifugal pump 25 with a manifold system. The pump 25 has carbon tip blades which assists in the life extension of the blades. The high energy mixing and the grinding action of the slurry unit 14 causes the slurry material to be sheared and therefore reduced in particle size. From the slurry tanks 14, the slurry is pumped over a second stage separation device 16 where the slurry passes over another set of screens that are sized accordingly to the well specification. The screens catch and separate any coarse fractions that escaped the grinding action from the slurry tank 14. The coarse fraction can be stockpiled or re-circulate through the waste processing system 1 of the present invention. The fine fractions that passes through the second stage separation device 16 are then fed into the 150 bbl mixing tank 18 which are equipped with agitators and centrifugal pumps 35. The second mixing tank 18 is used to regulate the control of all the slurry that will be injected into a wellbore. The second mix tank 18 also has a manifold system that enables us to feed the pumps 35 or re-circulate mud to any stage of the waste processing system 1. The manifold can be use to re-circulate the waste product to the 250 bbl mix tank 10. The processed slurrified fine particles are sent to the 400 bbl holding tank. The holding tank 20 is set up with four compartments 20a-d, each compartment being at 100 bbls. The holding tank 20 is set up with a suction manifold that is valved at each compartment. It is also manifolded to discharge into each compartment or back into the receiving line. The compartments of the holding tank 20 has an agitation system. The agitation system comprises a gear box, a motor and a plurality of blades. The 400 bbl holding tank 20 has at least two centrifugal pumps 35 that feed injection pumps 45 that are used for injecting the slurry into the wellbore. Obviously, numerous modifications and variations of the present invention are possible in light of the above teachings. It is therefore to be understood that with the scope of the claims appended hereto, the invention may be practical otherwise then as specifically disclosed herein.
	description	1. Field of the Invention The present invention relates to a defect repair apparatus for an EUV mask that repairs a defect in an EUV mask using a charged-particle beam. 2. Background Art In the related art, there is a technique of repairing a defect in a photo-mask employed for a lithographic technique using a focused ion beam. A lithographic technique using EUV (Extreme Ultra Violet) as an exposing source has been developed recently. A mask employed for EUV lithography is formed of a reflection layer made up of ultra-thin films having a multi-layer structure and an absorption layer of a pattern shape. A focused ion beam is irradiated to a defect in the pattern shape of the absorption layer to repair the defect by etching processing or deposition processing. An example is disclosed, for example, in JP-A-2009-210805. According to the disclosed technique as above, it becomes possible to repair a defect in the EUV mask. The technique in the related art, however, has a problem that anion beam irradiated to the reflection layer in the EUV mask causes damage to the multi-layer structure, which deteriorates reflectance markedly. The invention was devised in view of the foregoing and therefore has an object to provide a defect repair apparatus for an EUV mask that repairs a defect without markedly deteriorating reflectance of the reflection layer in the EUV mask. In order to provide the defect repair apparatus for an EUV mask as above, it is preferable to use an ion beam that causes less irradiation damage to the EUV mask. In addition, a considerable variance in amount of current of an ion beam during a defect repair causes excessive or insufficient processing. It is therefore necessary to irradiate an ion beam with a stable amount of current to the EUV mask. Further, because the EUV mask has a fine pattern, it is preferable to repair a defect with a high degree of accuracy. In view of the foregoing, the invention provides the following. A defect repair apparatus for an EUV mask according to one aspect of the invention includes: a gas field ion source that generates a hydrogen ion beam; an ion optical system that focuses the hydrogen ion beam onto the EUV mask; a sample stage on which to place the EUV mask; a detector that detects secondary charged-particles generated from the EUV mask; and an image forming unit that forms an observation image of the EUV mask on the basis of an output signal from the detector. When configured in this manner, an observation image can be obtained by irradiating a hydrogen ion beam to the EUV mask. It thus becomes possible to lessen damage to the multi-layer structure of ultra-thin films forming a reflection layer in the EUV mask. It is preferable that the defect repair apparatus for an EUV mask further includes a hydrogen gas supply source that supplies a hydrogen gas to the gas field ion source and a purifier that purifies the hydrogen gas. When configured in this manner, a high-purity hydrogen gas can be supplied to the gas field ion source. It thus becomes possible to irradiate a hydrogen ion beam with a stable amount of beam current. It is preferable that the defect repair apparatus for an EUV mask further includes an ion generation chamber in which to generate ions and an intermediate chamber provided between the ion generation chamber and a vacuum sample chamber. When configured in this manner, it becomes possible to reduce an inflow of an impurity gas from the vacuum sample chamber to the ion generation chamber. It is preferable that the defect repair apparatus for an EUV mask further includes a current measuring electrode that is provided between the gas field ion source and a focusing lens electrode and measures an amount of current of the hydrogen ion beam. When configured in this manner, it becomes possible to measure an amount of current of the hydrogen ion beam irradiated from the gas field ion source. Accordingly, it becomes possible to control ion beam irradiation conditions on the basis of the measured amount of current for the gas field ion source to irradiate a hydrogen ion beam with a stable amount of current. It is preferable for the defect repair apparatus for an EUV mask that the gas field ion source and the ion optical system irradiate the hydrogen ion beam having a beam diameter of 5 nm or less onto the EUV mask. Accordingly, it becomes possible to repair a fine pattern of the EUV mask with a high degree of accuracy. It is preferable for the defect repair apparatus for an EUV mask that an upper limit value of an amount of irradiation of the hydrogen ion beam onto the EUV mask is 4×1016 ions/cm2. When configured in this manner, it becomes possible to further lessen damage to the multi-layer structure of the ultra-thin films forming the reflection layer in the EUV mask. It is preferable that the defect repair apparatus for an EUV mask further includes a deposition gas supply system that supplies a deposition gas to the EUV mask. When configured in this manner, by supplying a deposition gas and irradiating a hydrogen ion beam to the EUV mask, a deposition film can be formed locally. It thus becomes possible to repair a defect portion of the absorption layer in the EUV mask with the deposition film. It is preferable that the defect repair apparatus for an EUV mask further includes an etching gas supply system that supplies an etching gas to the EUV mask. When configured in this manner, by supplying an etching gas and irradiating a hydrogen ion beam to the EUV mask, etching processing can be applied locally at a high speed. It thus becomes possible to repair a defect by effectively applying etching processing to a defect portion of the absorption layer in the EUV mask. A defect repair method for an EUV mask according to another aspect of the invention includes: obtaining an observation image by scanning and irradiating a hydrogen ion beam onto the EUV mask; setting a defect repair position from the observation image; and repairing a defect by irradiating the hydrogen ion beam to the defect repair position. When configured in this manner, an observation image can be obtained by scanning and irradiating a hydrogen ion beam to the EUV mask. It thus becomes possible to lessen damage to the multi-layer structure of the ultra-thin films forming the reflection layer in the EUV mask. It is preferable for the defect repair method for an EUV mask that an upper limit value of an amount of irradiation of the hydrogen ion beam is 4×1016 ions/cm2. When configured in this manner, it becomes possible to further lessen damage to the multi-layer structure of the ultra-thin films forming the reflection layer in the EUV mask. It is preferable for the defect repair method for an EUV mask that a deposition gas is supplied to the defect when the defect is repaired. When configured in this manner, by supplying a deposition gas and irradiating a hydrogen ion beam to the EUV mask, a deposition film can be formed locally. It thus becomes possible to repair a defect portion of the absorption layer in the EUV mask with the deposition film. It is preferable for the defect repair method for an EUV mask that an etching gas is supplied to the defect when the defect is repaired. When configured in this manner, by supplying an etching gas and irradiating a hydrogen ion beam to the EUV mask, etching processing can be applied locally at a high speed. It thus becomes possible to repair a defect by effectively applying etching processing to a defect portion of the absorption layer in the EUV mask. According to the defect repair apparatus for an EUV mask configured as above, it becomes possible to repair a defect in the EUV mask without markedly deteriorating reflectance of the reflection layer in the EUV mask. Hereinafter, embodiments of a defect repair apparatus for an EUV mask of the invention is explained. (1) Defect Repair Apparatus for EUV Mask As is shown in FIG. 1, a defect repair apparatus 100 for an EUV mask according to one embodiment includes an ion beam column 1 provided with an ion optical system having a focusing lens electrode 16 and an objective lens electrode 17 that focus ions generated in an ion source 12 onto a sample 3 placed within a vacuum sample chamber 11. The defect repair apparatus 100 for an EUV mask also includes a secondary electron detector 5, which is a secondary charged-particle detector that detects secondary electrons 4 generated when an ion beam 2 is irradiated to the sample 3 from the ion beam column 1. Herein, a secondary ion detector is used as the secondary charged-particle detector in a case where secondary ions generated from the sample 3 are detected. The defect repair apparatus 100 for an EUV mask also includes a gas supply system 6 that supplies a gas to the surface of the sample 3, a sample holder 7 that fixes the sample 3, and a sample stage 8 that moves the sample 3. The defect repair apparatus 100 for an EUV mask uses a reflected ion detector as the secondary charged-particle detector in a case where reflected ions generated from the sample 3 are detected. The defect repair apparatus 100 for an EUV mask further has an image forming section that includes an image forming unit 9 that forms an observation image from a scan signal of an ion beam 2 and a detection signal from the secondary electron detector 5 and a display unit 10 that displays an observation image thereon. (2) Ion Source The ion source 12 is a gas field ion source and, as is shown in FIG. 2, includes an ion generation chamber 21, an emitter 22, an extraction electrode 23, and a cooling device 24. The cooling device 24 is provided to a wall portion of the ion generation chamber 21. The emitter 22 of a needle shape is attached to the cooling device 24 on a surface facing the ion generation chamber 21. The cooling device 24 cools the emitter 22 with a cooling medium, such as liquid nitrogen and liquid helium, stored inside thereof. As the cooling device 24, a GM or pulse tube closed cycle refrigerator or a gas flow refrigerator is available. The extraction electrode 23 having an opening at a position opposing a tip end 22a of the emitter 22 is provided in the vicinity of an opening end of the ion generation chamber 21. The ion generation chamber 21 maintains the interior thereof in a high-vacuum state using an air exhauster. A hydrogen gas supply source 40 is connected to the ion generation chamber 21 via gas introduction tubes 43, 44, and 45, so that a slight amount of a hydrogen gas is supplied into the ion generation chamber 21. The emitter 22 is a member formed by coating a needle-shaped base material made of tungsten or molybdenum with precious metal, such as platinum, palladium, iridium, rhodium, and gold. The tip end 22a is of a pyramidal shape pointed sharply at the atomic level. Alternatively, the emitter 22 may be the one formed of a needle-shaped base material made of tungsten or molybdenum with the tip end 22a pointed sharply at the atomic level by introducing a nitrogen gas or an oxygen gas to the base material. The emitter 22 is maintained at a temperature as low as or lower than about 100 K by the cooling device 24 while the ion source 12 is in operation. A voltage is applied between the emitter 22 and the extraction electrode 23 by a power supply 27. When a voltage is applied between the emitter 22 and the extraction electrode 23, an extremely large electric field is developed at the sharply pointed tip end 22a while hydrogen molecules 25 are polarized and attracted to the emitter 22 and turn into hydrogen ions as they lose electrons through tunneling at a high field position of the tip end 22a. The hydrogen ions are repelled by the emitter 22 maintained at positive potential and come flying out toward the extraction electrode 23. Hydrogen ions 28 ejected to an ion optical system through the opening in the extraction electrode 23 form an ion beam 2. In the case of a gas field ion source, hydrogen ions include hydrogen molecular ions and hydrogen atomic ions and a ratio thereof varies with a voltage. Under normal use conditions, most of the hydrogen ions are hydrogen molecular ions. The tip end 22a of the emitter 22 is of an extremely pointed shape. Because hydrogen ions fly out from the tip end 22a, an energy distribution width of an ion beam 2 emitted from the ion source 12 is extremely narrow. For example, in comparison with a plasma gas ion source or a liquid metal ion source, an ion beam with a smaller beam diameter at higher intensity can be obtained. When an applied voltage to the emitter 22 is too large, constituent elements (tungsten and platinum) of the emitter 22 fly out toward the extraction electrode 23 together with hydrogen ions. Accordingly, a voltage to be applied to the emitter 22 during operation (while an ion beam is being emitted) is maintained to a voltage small enough for constituent elements of the emitter 22 not to fly out from the emitter 22. Meanwhile, the shape of the tip end 22a can be adjusted by taking advantage of the capability of engineering the constituent elements of the emitter 22 as described above. For example, the ion beam diameter can be increased by broadening a region to ionize a gas by intentionally removing elements positioned at the leading edge of the tip end 22a. In addition, by heating the emitter 22, precious metal elements on the surface can be repositioned while preventing the precious metal elements from flying out. It thus becomes possible to restore the sharply pointed shape of the tip end 22a that has become dull with use. Incidentally, when water molecules are contained in a hydrogen gas supplied to the ion generation chamber 21, water adheres to the emitter 22 and forms a protrusion. Hydrogen ions are then released in a direction different from the optical axis of an ion beam 2. Because adhesion of water molecules occurs randomly, an amount of current of an ion beam 2 released in the optical axis direction may possibly vary considerably. In order to avoid such an inconvenience, a hydrogen gas to be supplied to the ion generation chamber 21 is purified. Metal tubes are used as the gas introduction tubes 43, 44, and 45. It is particularly preferable to use SUS-EP tubes having a surface roughness made finer by electropolishing. By pre-heating the gas introduction tubes 43, 44, and 45 to several hundred degrees centigrade, adhesion of water onto the inner surface of the tubes can be lessened. Also, purifiers to purify a hydrogen gas supplied from the hydrogen gas supply source 40 are provided. A first purifier 41 purifies a hydrogen gas by allowing an impurity gas to be absorbed into a getter material made of more than one type of activated metal or to pass through a heated palladium thin film. A second purifier 42 eliminates an impurity with a cold trap using liquid nitrogen. Consequently, a high-purity hydrogen gas at 9N purity (99.9999999%) or above can be supplied. Either the first purifier or the second purifier alone may be used as the purifiers. Also, the purifiers may be incorporated into the hydrogen gas supply source 40. Further, in order to reduce an inflow of an impurity gas from the vacuum sample chamber 11 to the ion generation chamber 21, an intermediate chamber 13 in a vacuum state is provided inside the ion beam column 1. The interior of the intermediate chamber 13 is evacuated by a vacuum pump 14 different from the air evacuator used to evacuate the ion generation chamber 12. An ion beam 2 generated in the ion generation chamber 21 passes through a small-diameter hole between the vacuum chambers to be irradiated to the vacuum sample chamber 11. A hole 111 is provided between the ion generation chamber 21 and the intermediate chamber 13 and a hole 112 is provided between the intermediate chamber 13 and the vacuum sample chamber 11. Herein, it may be configured in such a manner that the intermediate chamber 13 accommodates the objective lens electrode 17 and the hole 112 is positioned closer to the vacuum sample chamber 11 than to the objective lens electrode 17. In particular, in a case where a deposition material gas or an etching gas for a deposition film is used for a defect repair in the vacuum sample chamber 11, an inflow of an impurity gas in the deposition material gas or the etching gas into the ion generation chamber 21 can be reduced. (3) Ion Optical System The ion optical system includes, sequentially from the ion source 12 to the vacuum sample chamber 11, the focusing lens electrode 16 that converges an ion beam 2 and the objective lens electrode 17 that focuses the ion beam 2 onto the sample 3. With the ion beam column 1 configured as above, because a source size can be set to 1 nm or smaller and energy dispersion of an ion beam 2 can be set to 1 eV or lower, a beam diameter can be reduced to 5 nm or smaller. Although it is not shown in the drawing, the ion beam column 1 may be provided with a mass filter, such as an E×B mass filter, to discriminate the atomic number of ions. (4) Current Amount Measurement The ion beam column 1 includes a current measuring electrode 18 to measure an amount of current of an ion beam 2 between the ion source 12 and the focusing lens electrode 16. An ammeter 19 connected to the current measuring electrode 18 measures an amount of current of an ion beam 2 irradiated to the current measuring electrode 18. The extraction electrode 23 in the ion source 12 is controlled so that an amount of current measured by the ammeter 19 stays constant. It thus becomes possible to irradiate an ion beam 2 with a stable amount of current to the sample 3. (5) Gas Supply System The gas supply system 6 is configured to supply a deposition material gas of a deposition film (for example, a carbon-based gas, such as phenanthrene and naphthalene, a metal compound gas containing metal, such as platinum and tungsten) onto the surface of the sample 3 through a gas nozzle from a deposition material container. In a case where etching processing is applied, the gas supply system 6 is capable of supplying an etching gas (for example, xenon fluoride, chlorine, iodine, chlorine trifluoride, fluorine monoxide, and water) through the gas nozzle from the deposition material container. (6) EUV Mask An EUV mask used as the sample 3 is formed of, as is shown in FIG. 3, a reflection layer 33 of an Mo/Si multi-layer structure, a buffer layer 32, an absorber 31 (pattern shape), which are provided sequentially on a glass substrate 34. According to EUV lithography, EUV light is irradiated on the EUV mask and a mask pattern is transferred using reflected light. When there is a defect in the pattern shape of the absorber 31 of the EUV mask, the pattern shape together with the defect is transferred. In order to avoid such an inconvenience, it is necessary to repair the defect. Regarding the pattern dimension of the EUV mask, for example, in a case of a process corresponding to a 22-nm node, a half pitch of a line and a space is 88 nm and required defect repair accuracy is about 2 nm or smaller for three sigmas. (7) Ion Beam Irradiation Damage on EUV Mask The inventors of the present application conducted a simulation and an experiment for ion beam irradiation damage on an EUV mask. In the experiment, a damaged state of the EUV mask by irradiation of an ion beam to the Mo/Si multi-layer structure, which is the reflection layer, was examined. Helium ion beams with amounts of beam irradiation of 4×1015 ions/cm2, 4×1016 ions/cm2, and 4×1017 ions/cm2 were irradiated at an acceleration voltage of 30 kV and sectional TEM (Transmission Electron Microscopy) images of the Mo/Si multi-layer structure after irradiation were obtained. Then, it was found that mixing had occurred in the Mo/Si multi-layer structure to a depth of 280 nm from the sample surface in the EUV masks irradiated by helium ions with amounts of beam irradiation of 4×1016 ions/cm2 and 4×1017 ions/cm2. A simulation in the process of ion implantation by the Monte Carlo method was also conducted and the ion implantation depth was in agreement with the experiment result. Reflectance of EUV masks for EUV light after irradiation of ion beams was measured. Then, it was found that reflectance was deteriorated markedly by irradiation of ion beams with amounts of beam irradiation of 4×1016 ions/cm2 and 4×1017 ions/cm2. In view of the foregoing, it can be said that mixing occurs in the multi-layer structure by irradiating an ion beam onto the reflection layer and reflectance of the EUV mask for EUV light is consequently deteriorated. Also, damage becomes larger as an amount of beam irradiation increases. With a helium ion beam, it was found that deterioration in reflectance is small when an amount of beam irradiation was 4×1015 ions/cm2, which is, however, insufficient to repair a defect portion. Given these circumstances, the inventors of the application paid attention to hydrogen ions having a smaller atomic number than helium ions. A simulation in the process of ion implantation by the Monte Carlo method was conducted and a comparison was made as to the numbers of repelled atoms within the sample when incident ions were implanted. Then, it was found that when incident ions were hydrogen ions, the number of repelled ions was about 1/10 of that when incident ions were helium ions. On the basis of this finding, the inventors achieved an idea that a hydrogen ion beam is effective as a beam used to repair a defect in an EUV mask. Hence, the inventors of the application conducted an experiment to check a relation between an amount of irradiation of a hydrogen ion beam and a damaged state of the reflection layer in the EUV mask. Hydrogen ion beams with amounts of beam irradiation of 4×1016 ions/cm2 and 4×1017 ions/cm2 were irradiated at an acceleration voltage of 30 kV and sectional TEM (Transmission Electron Microscopy) images of the Mo/Si multi-layer structure after irradiation were obtained. Then, it was found that a damaged state in the sectional TEM image in the case of an amount of beam irradiation of 4×1016 ions/cm2 was substantially equal to a damaged state in the case of a helium ion beam with an amount of beam irradiation of 4×1015 ions/cm2. In other words, a hydrogen ion beam with an amount of beam irradiation of 4×1016 ions/cm2 is practical because even when it is irradiated to the reflection layer, deterioration in reflectance of the reflection layer is small and an amount of irradiation is sufficient to repair a defect portion. Also, it was found that a region where mixing occurred was 140 nm deep from the sample surface. This result is in agreement with the result of the simulation in the process of ion implantation by the Monte Carlo method. In view of the foregoing, it can be said that it becomes possible to repair a defect in an EUV mask while lessening damage to the EUV mask by using a hydrogen ion beam. First Embodiment One embodiment of the application will now be described along the flowchart of FIG. 4. A defect position in an EUV mask, which is the sample 3 placed on the sample holder 7, is moved onto the sample stage 8 to be in an irradiation region of an ion beam 2. A hydrogen ion beam is scanned and irradiated to the EUV mask from the ion beam column 1 and secondary electrons 4 generated from the EUV mask are detected by the secondary electron detector 5. An observation image of the EUV mask is obtained in the image forming unit 9 from a scan signal of the hydrogen ion beam and a detection signal from the secondary electron detector 5 (observation image obtaining step S1). Subsequently, the obtained observation image is displayed on the display unit 10 and a repair position setting to set the ion beam irradiation region on a defect portion is carried out (repair position setting step S2). The ion beam irradiation position is thus determined. Subsequently, in a case where the defect is a redundant defect of the absorber pattern, xenon fluoride as an etching gas is supplied to the surface of the sample 3 from the gas supply system 6 and defect repair processing is applied by irradiating a hydrogen ion beam to the ion beam irradiation region (defect repair step S3). The processing is ended when the redundant defect portion is etched away. In a case where the defect is a missing defect of the absorber pattern, a phenanthrene gas as a deposition gas is supplied to the surface of the sample 3 from the gas supply system 6 and the defect repair processing is applied by irradiating a hydrogen ion beam to the ion beam irradiation region. The defect is repaired as a deposition film is deposited on the defect portion. The repair result is confirmed (S4) and the flow ends when the repair is completed. In a case where the repair is not completed, the flow returns to the observation image obtaining step S1 to repair the defect portion again. By using hydrogen ions as the ion seed of an ion beam, ion beam irradiation damage on the reflection layer caused by irradiation of an ion beam when repairing a defect portion can be lessened more considerably than in a case where helium ions are used. It thus becomes possible to lessen deterioration in reflectance of the reflection layer for EUV light. Second Embodiment An embodiment where the upper limit value of an amount of hydrogen ion beam irradiation to an EUV mask is set to 4×1016 ions/cm2 will be described. FIG. 5 is a surface view of a part of an EUV mask and the pattern of the absorber 31 is provided on the reflection layer 33. An ion beam 2 is scanned and irradiated on an observation region 52 including a defect 51 in the EUV mask. Secondary electrons 4 being generated are detected by the secondary electron detector 5 and an observation image of the observation region 52 is obtained. A repair position is set on the obtained observation image and the defect is repaired. In a case where the repair is not completed, an ion beam 2 is scanned and irradiated again on the observation region 52 to obtain an observation image. Then, a repair position for additional processing is set on the observation image. The upper limit value of an amount of hydrogen ion beam irradiation referred to herein means the upper limit value of an amount of ions irradiated onto a given irradiation region. In the repair processing above, it is controlled so that a total amount of ions irradiated onto the defect 51 does not reach the upper limit value. The defect 51 is of a dimension ranging from several nm to one micron meter and a processing region is set depending on the size of a defect. A current of 5 pA or below is used as an ion beam current and the ion beam current is measured in advance. A dose of irradiation per unit area is calculated on the basis of the processing region and the current to indicate the maximum time over which irradiation is allowed. This enables an operator to perform a repair operation by paying attention to the upper limit value of an amount of irradiation. In a case where an amount of irradiation has reached the upper limit value, it is possible to perform control to inhibit irradiation of an ion beam 2 onto the same observation region. It thus becomes possible to prevent an ion beam 2 being irradiated onto the EUV mask with an amount of irradiation exceeding the upper limit value.
	abstract	The present disclosure relates to a device for conversion of one type of energy into another type of energy. Specifically, the device converts radiation energy into electrical energy.
041522044	summary	The present invention relates to a device for controlling the power of and turning off a core reactor with a reflector of graphite, which reflector contains a filling or pile of ball-shaped fuel and/or breeder elements, while the component of the device which contains absorbing material is, for controlling the power of the core reactor and/or for turning off the reactor, by that part of the reflector wall which is located above said pile or filling, movable up to a predetermined depth into the chamber surrounded by the reflector. During this movement, and due to the absorbing material, the neutrons which become free during the fission process are absorbed so that the reactivity of the reactor is lowered. With the heretofore known core reactors with a pile or filling of ball-shaped fuel and/or breeder elements, that component of the device for power control and turning off the core reactor, which component contains absorbing material, consists of cylindrical rods which for controlling the reactor are moved into the filling or into the reflector surrounding the filling. By varying the penetration depth of said rods, the reactivity of the core reactor is changed and, in view of the change in reactivity, the power output of the core reactor is increased or reduced. With core reactors having an electrical power output of up to 300 MW, a device with absorber rods moved into the reflector surrounding the pike or filling was preferred. In order to make sure that also with possibly occurring displacements of the graphite blocks of the reflector, the ability of the device to function will be maintained, the rods were made of a plurality of elements movable relative to each other. In view of the rods moved into the reflector, the neutron flow in the reflector and into the outer zones of the filling adjacent the reflector were lowered whereby the reactivity of the core reactor was reduced. This design, however, had the drawback that the thermal neutron flow was displaced into the interior region of the reactor core. As a result thereof, the temperature in the interior range of the filling was considerably higher than in the marginal zone of the filling so that the cooling gas flowing through the axis-near region of the filling reached higher temperatures than that portion of the cooling gas which flowed through the marginal zone of the filling. This is disadvantageous because in the inserts which follow the filling in the direction of flow of the cooling gas, for instance, in the heat exchanger, certain tensions occurred due to the temperature drop, and as a result these inserts were continuously subjected to temperature change loads, also in view of the continuously changing directions of the flow lines of the cooling gas. With larger core reactors with an electrical output of, for instance, 1000 MW, a control by means of absorber rods movable into the reflector surrounding the filling was no longer possible because the decrease in the neutron flow obtained by moving in the absorber rods no longer sufficed to control the power of the core reactor. In order to be able at any desired time to control a load change from 100% to 40% and from 40% again to 100% of the output of the core reactor, it has been suggested with larger core reactors to provide an arrangement which is movable into the pile or filling and within the filling or pile in a predetermined manner, and which contains absorbing material. This heretofore known arrangement comprised a plurality of cylindrical rods, which arrangement was movable into the filling or pile through the reactor wall above the filling or pile. While this arrangement made possible a control of the output of even larger core reactors of the above mentioned type, it has proved disadvantageous that in this connection a pressure was exerted by the absorber rods upon the fuel and/or breeder elements adjacent to said absorber rods. This pressure became higher the greater necessarily the number of the rods was and the deeper the rods were moved into the pile or filling. Due to the compression of the filling as it occurred during the movement of the absorber rods in the filling, it was necessary to exert an increased pressure in order to bring the absorber rods to the penetration depth necessary for the control. The maximum pressure exerted in this connection by the fuel and/or breeder elements frequently approached the boundaries of the mechanical loadability of the fuel and/or breeder elements. Inasmuch as it was very difficult to foresee the effect of the compression in the filling as brought about by the displacement of the absorber rods within the filling, upon the moving of the absorber rods into the filling it was necessary so to design the driving units of the absorber rods that the absorber rods could be moved upwardly and downwardly in the filling even if the filling was compressed to a high extent. Such an arrangement entailed, of course, high costs. A further drawback of the control of the output of the core reactor by means of absorber rods movable into the filling consisted in that fuel and/or breeder element balls in the vicinity of which, and for a longer period of time, and absorber rod moved into the filling was located and which fuel and/or breeder element balls during this time period were exposed only to a low thermal neutron flow, were exposed to a high thermal neutron flow as soon as the absorber rods were pulled out of the filling. The disadvantageous result was that very high output peaks occurred in the fuel and/or breeder elements. As further drawback in connection with the control of a core reactor by means of absorber rods moved into the graphite of the reflector or by means of an absorber moved into the filling, consists is that the efficiency of the absorber rods was limited in view of the occurring alternating effect with their direct surrounding. In this connection, it should be noted that due to the absorption of the thermal neutrons into the absorber rods, the flow of neutrons in the immediate vicinity of the absorber rods is lowered so that the lowering of the neutron flow of the core reactor obtainable by the individual absorber rods is reduced. This entailed the disadvantageous result that the arrangement for controlling the output of the core reactor required a great number of absorber rods, or the absorber rods for controlling the output had to be moved into the filling or pile. Therefore, it is an object of the present invention to provide a simple and thus economically produceable device for controlling the power output of a core reactor and for turning off the core reactor, which will overcome the above mentioned drawbacks. It is another object of this invention to provide a control device as set forth in the preceding paragraph, in which the heretofore encountered drawbacks in connection with the controlling of the power output of a core reactor and the turning off a core reactor will be avoided even if the electrical output of the core reactor exceeds 300 MW. The above outlined problem has, according to the present invention, been solved with a core reactor according to the present invention, by composing that component of the device which contains the absorbing material, of a part for turning off the core reactor and of a part of controlling the output of the core reactor while that part which serves for turning off the core reactor is movable into the filling and while that part of the device which serves for controlling the output is movable within the wall of that portion of the reflector which is located above the filling and is displaceable in the hollow chamber confined by the surface of the filling and by that portion of the reflector which is located above the filling. The penetration depth of that portion which serves for controlling the output power is confined from below by the surface of the filling. The invention is based on the finding that due to that part of the device which is moved into the wall of the portion of the reflector located above the filling and due to the portion of the device moved into the hollow chamber above the filling, the neutron flow in the upper part of the reflector and in said hollow chamber and thus in the pile of the fuel elements is reduced. In this connection, it has been found that the neutron flow in the reactor core and thus the output of the reactor are lowered all the more, the more absorbing material acting as neutron absorber is introduced into said hollow chamber. Surprisingly, it has been found that the efficiency of that component of the device which contains that portion of the absorbing material which is introduced into said hollow chamber for controlling the core reactor, and which component is intended for the power output control and for turning off the core reactor (in contrast to heretofore known core reactors with absorber rods moved into the reflector or into the filling) will not be reduced by lowering the neutron flow in the immediate vicinity of that portion of the device which is moved into said hollow chamber. Rather--due to the very long free length of the path of the neutron is said hollow chamber--the neutron flow due to the portion of the device introduced into the hollow chamber for the power output control and the turning off of the core reactor is reduced in the entire vicinity of the hollow chamber, in other words, also in the freshly added fuel and/or breeder elements in the uppermost layers of the filling. A considerable advantage of the device according to the present invention thus consists in that that portion of the device which is intended for controlling the output of the core reactor is moved only in the reflector wall located above the hollow chamber, and in the hollow chamber itself. During the power output control, therefore, no force has to be overcome which acts against the direction of movement of the pertaining part of the device for controlling the power output and for turning off the device. This is in contrast to the heretofore known core reactors in which the power output control is effected by means of absorber rods moved into the filling. It is also advantageous that the component which contains the absorber material and serves for controlling the power output is, according to the invention, considerably shorter than is the case of comparable devices of heretofore known core reactors. This entails the further advantage that for driving that portion of the device which serves for controlling the power smaller driving units can be employed than was the case with heretofore known core reactors. It has also proved extremely advantageous that due to that portion of the device according to the invention which is inserted into the hollow chamber for controlling the power output, the neutron flow in the vicinity of the hollow chamber and thus also in the fuel and/or breeder elements of the upper layer of the filling and -- even though to a minor extent in the lower layers of the filling is uniformly lowered without local output peaks occurring in the filling, which peaks occur with heretofore known core reactors which are controlled by means of absorber rods introduced into the filling. This also favorably affects the temperature distribution in the cooling gas as well as the life span of the fuel and/or breeder elements. When the core reactor with its device according to the invention is operated in such a way that the fuel and/or breeder elements pass through the filling only once, it has been found that that portion of the device according to the invention becomes particularly effective which for purposes of controlling the core reactor is moved into the hollow chamber. It is in this way that the neutron flow in the hollow chamber and in the vicinity of the hollow chamber is reduced. In this way, the axial neutron flow profile is so deformed in the core of the reactor that the absorbing effect of the fission products in the burned-off fuel and/or breeder elements increases in the lower core range. As has been proved, the efficiency of that portion of the device according to the invention which contains the absorbing material is increased by from 20 to 40%. According to a further advantageous development of the device according to the invention, that portion of the device which is intended for controlling the output of the core reactor is composed of a plurality of plates containing absorbing material. The design of the plates, the longest extension of which is expediently in the direction of the axis of the reactor core is limited only by its mechanical bending strength which is anyhow subjected only to low stresses because the plates are not moved into the filling. Since plates of this type have a considerably larger surface than cylindrical absorber rods, it will be appreciated that simultaneously the surface containing absorbed material is greatly enlarged. This brings about, with the same effect, a considerable saving in material. It has also proved expedient when that portion of the device which serves for turning off a core reactor, for controlling the output of the core reactor is displaceable also within the wall of the part of the reflector located above the filling and within the hollow chamber limited by the surface of the filling and that portion of the reflector which is located above the filling. Due to the fact that that portion of the device according to the invention which serves for turning off the core reactor is during the operation of the reactor, also employed for controlling the power output, it is possible further to simplify that portion of the device which serves the control of the power output. By way of example, two core reactors of the same output and of the same geometric dimensions will not be compared with each other. One of said core reactors comprises a device for controlling the power output and for turning off the reactor according to the present invention, whereas the device for power control of the other reactor is composed of absorber rods which are moved into the filling. Both core reactors have a total output of 500 MW.sub.th. The height of the core amounts to approximately 600 cm. The radius is precisely 240 cm. The height of the hollow chamber is 100 cm. The fuel elements consist of covered particles of UO.sub.2 which are arranged in a graphite ball, said balls having an outer radius of 3 cm. The heavy metal content amounts per ball to 10.7 g per each ball of 6.5% enriched uranium. The pouring density of the balls in the core amounts to 0.61. As cooling gas, helium of 40 atmospheres is employed. The thickness of the upper reflector amounts to 100 cm., whereas the thickness of the lower reflector amounts to 150 cm. The thickness of the mantle reflector is 100 cm. The graphite density is approximately 1.6 grams per cubic cm. The absorbing material containing part of the device for controlling the output of the reactor for turning off the reactor comprises with both reactors 19 absorber rods each having a diameter of 13 cm. While one rod is arranged centrally in the reactor core and each 6 rods are arranged along circles respectively having radii of 100 cm, 130 cm, and 175 cm around the central rod. Experience has shown that for controlling the core reactor at a load change operation from 100% to 40% and from 40% to 100% makes necessary a reactivity change of the reactor by 2.4%. This change is realized with the heretofore known core reactor by changing the penetration depth of the 19 rods in the filling or pile by approximately 60 cm. With the core reactor having a device according to the present invention, the reactivity is by moving the rods into the wall of the upper reflector decreased by approximately 0.7%, and by moving the rods into the hollow chamber is decreased by an additional 1.8%, which means is lowered by a total of 2.5%. Thus, with a portion of the device comprising cylindrical absorber rods and serving the control of the output, it is possible when moving said rods into the upper reflector and into the hollow chamber, to realize a sufficiently high change of the neutron flow in order to be able to carry out a load change from 100% to 40% and from 40% to 100% at any desired time. Such load change is with the core reactor with a simplified embodiment of the device according to the invention but with otherwise equal geometric dimensions and output of the reactor likewise obtainable at any time. If, namely, instead of 19 cylindrical absorber rods, 8 plates with a cross section of 8 times 40 cm are arranged circularly around the axis of the reactor core with a radius of 150 cm, it will be possible by moving said 8 plates into the wall of the upper reflector and into the hollow chamber likewise to realize a reactivity change by 2.5%.
	abstract	A jet pump slip joint repair assembly includes at least one clamp and a bushing configured to be inserted in a bore of a diffuser and to surround a portion of an inlet mixer. The clamp includes a gripping surface and a gripping collar. The bushing includes a generally cylindrical sidewall, the sidewall configured to surround the portion of the inlet mixer, a grooved flange on an upper surface of the sidewall, at least one cutout between adjacent portions of the grooved flange, and a groove on an inner, bottom surface of the sidewall. The assembly also includes a seal in the groove. The seal is flexible and formed of a metallic material. The seal is configured to be compressed when the at least one clamp engages the bushing.
	description	The present application is a continuation application of International Application No. PCT/JP2018/042360, filed on Nov. 15, 2018, the entire contents of which are hereby incorporated by reference. The present disclosure relates to an extreme ultraviolet light generation apparatus and an electronic device manufacturing method. Recently, miniaturization of a transfer pattern in optical lithography of a semiconductor process has been rapidly proceeding along with miniaturization of the semiconductor process. In the next generation, microfabrication at 20 nm or less will be required. Therefore, it is expected to develop an exposure apparatus that combines an apparatus for generating extreme ultraviolet (EUV) light having a wavelength of about 13 nm with a reduced projection reflection optical system. As an EUV light generation apparatus, three types of apparatuses have been proposed: a laser produced plasma (LPP) type apparatus using plasma generated by irradiating a target substance with laser light, a discharge produced plasma (DPP) type apparatus using plasma generated by discharge, and a synchrotron radiation (SR) type apparatus using synchrotron radiation light. Patent Document 1: US Patent Application Publication No. 2018/0224748 Patent Document 2: US Patent Application Publication No. 2013/0161540 An extreme ultraviolet light generation apparatus according to an aspect of the present disclosure includes a chamber configured to cause a target substance to be turned into plasma with laser light radiated to a plasma generation region at an internal space of the chamber, a light concentrating mirror configured to concentrate extreme ultraviolet light generated by the turning of the target substance into plasma, a gas supply unit configured to supply gas into the chamber, a magnetic field generation unit configured to generate a magnetic field including a magnetic field axis that crosses a light path of the extreme ultraviolet light, a first exhaust port arranged at a position through which the magnetic field axis passes in the chamber, a second exhaust port arranged at a position opposite to the light concentrating mirror in the chamber with reference to a plane passing through the first exhaust port and being perpendicular to an optical axis of the light concentrating mirror, and a gas exhaust amount adjustment unit configured to adjust a ratio between an exhaust amount of first exhaust gas exhausted from the first exhaust port and an exhaust amount of second exhaust gas exhausted from the second exhaust port. An electronic device manufacturing method according to an aspect of the present disclosure includes generating extreme ultraviolet light using an extreme ultraviolet light generation apparatus, emitting the extreme ultraviolet light to an exposure apparatus, and exposing a photosensitive substrate to the extreme ultraviolet light in the exposure apparatus to produce an electronic device. Here, the extreme ultraviolet light generation apparatus includes a chamber configured to cause a target substance to be turned into plasma with laser light radiated to a plasma generation region at an internal space of the chamber, a light concentrating mirror configured to concentrate the extreme ultraviolet light generated by the turning of the target substance into plasma, a gas supply unit configured to supply gas into the chamber, a magnetic field generation unit configured to generate a magnetic field including a magnetic field axis that crosses a light path of the extreme ultraviolet light, a first exhaust port arranged at a position through which the magnetic field axis passes in the chamber, a second exhaust port arranged at a position opposite to the light concentrating mirror in the chamber with reference to a plane passing through the first exhaust port and being perpendicular to an optical axis of the light concentrating mirror, and a gas exhaust amount adjustment unit configured to adjust a ratio between an exhaust amount of first exhaust gas exhausted from the first exhaust port and an exhaust amount of second exhaust gas exhausted from the second exhaust port. 1. Overview 2. Description of electronic device manufacturing apparatus 3. Description of extreme ultraviolet light generation apparatus of comparative example 3.1 Configuration 3.2 Operation 3.3 Problem4. Description of extreme ultraviolet light generation apparatus of first embodiment 4.1 Configuration 4.2 Operation 4.3 Effect5. Description of extreme ultraviolet light generation apparatus of second embodiment 5.1 Configuration 5.2 Operation 5.3 Effect6. Description of extreme ultraviolet light generation apparatus of third embodiment 6.1 Configuration 6.2 Operation 6.3 Effect7. Description of extreme ultraviolet light generation apparatus of fourth embodiment 7.1 Configuration 7.2 Operation 7.3 Effect Hereinafter, embodiments of the present disclosure will be described in detail with reference to the drawings. The embodiments described below show some examples of the present disclosure and do not limit the contents of the present disclosure. Also, all configurations and operation described in the embodiments are not necessarily essential as configurations and operation of the present disclosure. Here, the same components are denoted by the same reference numerals, and duplicate description thereof is omitted. 1. Overview Embodiments of the present disclosure relate to an extreme ultraviolet light generation apparatus generating light having a wavelength of extreme ultraviolet (EUV) light, and an electronic device manufacturing apparatus. In the following of the present specification, extreme ultraviolet light is referred to as EUV light in some cases. 2. Description of Electronic Device Manufacturing Apparatus As shown in FIG. 1, an electronic device manufacturing apparatus includes an EUV light generation apparatus 100 and an exposure apparatus 200. The exposure apparatus 200 includes a mask irradiation unit 210 including a plurality of mirrors 211, 212 and a workpiece irradiation unit 220 including a plurality of mirrors 221, 222. The mask irradiation unit 210 irradiates a mask pattern on a mask table MT through a reflection optical system with EUV light 101 incident from the EUV light generation apparatus 100. The workpiece irradiation unit 220 images the EUV light 101 reflected by the mask table MT onto a workpiece (not shown) disposed on the workpiece table WT through a reflection optical system. The workpiece is a photosensitive substrate such as a semiconductor wafer on which photoresist is applied. The exposure apparatus 200 synchronously translates the mask table MT and the workpiece table WT to expose the workpiece to the EUV light 101 reflecting the mask pattern. Through the exposure process as described above, a device pattern is transferred onto the semiconductor wafer, thereby a semiconductor device can be manufactured. 3. Description of Extreme Ultraviolet Light Generation Apparatus of Comparative Example 3.1 Configuration The EUV light generation apparatus of a comparative example will be described. FIG. 2 is a schematic view showing a schematic configuration example of an entire EUV light generation apparatus of the present example. As shown in FIG. 2, a laser device 30 is connected to the EUV light generation apparatus 100 of the present embodiment. The EUV light generation apparatus 100 of the present embodiment includes a chamber 10, a control unit 20, and a laser light delivery optical system 35. The chamber 10 is a sealable container. The chamber 10 is continuously provided with a sub-chamber 15, and a target supply unit 40 is provided in the sub-chamber 15. The target supply unit 40 includes a tank 41 and a nozzle 42. The target supply unit 40 supplies a droplet DL to the internal space of the chamber 10 and is attached, for example, to penetrate through a wall of the sub-chamber 15. The droplet DL, which is also called a target, is supplied from the target supply unit 40. The tank 41 stores therein a target substance which becomes the droplet DL. The target substance may include, but is not limited to, any one of tin, terbium, gadolinium, lithium, and xenon, or a combination of any two or more thereof. The inside of the tank 41 communicates, through a pipe, with a pressure adjuster 43 adjusting gas pressure. Further, a heater 44 is attached to the tank 41. The heater 44 heats the tank 41 with current applied from a heater power source 45. Through the heating, the target substance in the tank 41 melts. The pressure adjuster 43 and the heater power source 45 are connected to the control unit 20. The nozzle 42 is attached to the tank 41 and discharges the target substance. A piezoelectric element 46 is attached to the nozzle 42. The piezoelectric element 46 is connected to a piezoelectric power source 47 and is driven by voltage applied from the piezoelectric power source 47. The piezoelectric power source 47 is connected to the control unit 20. The target substance discharged from the nozzle 42 is formed into the droplet DL through operation of the piezoelectric element 46. The chamber 10 is provided with a target collector 14. The target collector 14 collects unnecessary droplets DL. At least one through hole is formed in a wall of the chamber 10. The through hole is blocked by a window 12 through which pulse laser light 301 emitted from the laser device 30 passes. Further, a laser light concentrating optical system 13 is located in the chamber 10. The laser light concentrating optical system 13 includes a laser light concentrating mirror 13A and a high reflection mirror 13B. The laser light concentrating mirror 13A reflects and concentrates the laser light 301 passing through the window 12. The high reflection mirror 13B reflects light concentrated by the laser light concentrating mirror 13A. Positions of the laser light concentrating mirror 13A and the high reflection mirror 13B are adjusted by a laser light manipulator 13C so that a laser light concentrating position at the internal space of the chamber 10 coincides with a position specified by the control unit 20. For example, an EUV light concentrating mirror 50 having a spheroidal reflection surface is disposed inside the chamber 10. The EUV light concentrating mirror 50 has first and second focal points. The EUV light concentrating mirror 50 may, for example, be disposed so that the first focal point is located in a plasma generation region AR and the second focal point is located at an intermediate focal point IF. A through hole is formed at the center of the EUV light concentrating mirror 50, and the pulse laser light 301 passes through the through hole. Further, the EUV light generation apparatus 100 includes a connection portion 19 providing communication between the internal space of the chamber 10 and the internal space of the exposure apparatus 200. A wall in which an aperture is formed is arranged in the connection portion 19. The wall is preferably disposed such that the aperture is located at the second focal point position of the EUV light concentrating mirror 50. Further, the EUV light generation apparatus 100 includes a pressure sensor 26. The pressure sensor 26 measures pressure at the internal space of the chamber 10. Further, the EUV light generation apparatus 100 includes a target sensor 27 attached to the chamber 10. The target sensor 27 has, for example, an imaging function, and detects the presence, trajectory, position, velocity, and the like of the droplet DL. The pressure sensor 26 and the target sensor 27 are connected to the control unit 20. The laser device 30 includes a master oscillator being a light source to perform burst operation. The master oscillator emits the pulse laser light 301 in a burst-on duration. The master oscillator is, for example, a laser device configured to emit the laser light by exciting, through electric discharge, gas as mixture of carbon dioxide gas with helium, nitrogen, or the like. Alternatively, the master oscillator may be a quantum cascade laser device. The master oscillator emits the pulse laser light 301 by a Q switch system. The master oscillator may include an optical switch, a polarizer, and the like. In the burst operation, the continuous pulse laser light is emitted at a predetermined repetition frequency in the burst-on duration and the emission of the laser light 301 is stopped in a burst-off duration. The travel direction of the laser light 301 emitted from the laser device 30 is adjusted by the laser light delivery optical system 35. The laser light delivery optical system 35 includes a plurality of mirrors 35A, 35B for adjusting the travel direction of the laser light 301, and a position of at least any of the mirrors 35A, 35B is adjusted by an actuator (not shown). Owing to that the position of at least any of the mirrors 35A, 35B is adjusted, the laser light 301 can appropriately propagate into the chamber 10 through the window 12. The control unit 20 includes a computer having a central processing unit (CPU) and the like. The control unit 20 controls the entire EUV light generation apparatus 100 and also controls the laser device 30. The control unit 20 receives a signal related to the pressure at the internal space of the chamber 10, which is measured by the pressure sensor 26, a signal related to image data of the droplet DL captured by the target sensor 27, a burst signal from the exposure apparatus 200, and the like. The control unit 20 is configured to process the image data and the like, and to control, for example, timing at which the droplet DL is output, an output direction of the droplet DL, and the like. Such various kinds of control described above are merely exemplary, and other control is added as described later. Next, the configuration of the chamber 10 will be described in more detail. FIG. 3 is a schematic view showing a schematic configuration of a part including the chamber 10 of the EUV light generation apparatus 100 of the comparative example. In FIG. 3, the laser light concentrating optical system 13 is omitted. As shown in FIGS. 2 and 3, the chamber 10 is provided with a first gas supply unit 72 and a second gas supply unit 73 for supplying etching gas into the internal space of the chamber 10. The first gas supply unit 72 and the second gas supply unit 73 are connected to the gas supply tank 74 which supplies the etching gas through pipes. When the target substance is tin, the etching gas is, for example, balancing gas having a hydrogen gas concentration of about 3%. The balance gas may include nitrogen (N2) gas or argon (Ar) gas. Hereinafter, description will be provided assuming that the target substance is tin and the etching gas contains hydrogen. The first gas supply unit 72 is adjusted so that the etching gas supplied into the chamber 10 flows along a reflection surface from the outer periphery of the EUV light concentrating mirror 50. Further, the second gas supply unit 73 has a truncated conical cylindrical shape, and is called a cone in some cases. A gas supply port of the second gas supply unit 73 is inserted into the through hole formed in the EUV concentrating mirror 50, and the second gas supply unit 73 supplies the etching gas from the through hole in a direction away from the EUV light concentrating mirror 50. Further, the laser light 301 passes through the through hole of the EUV light concentrating mirror 50 through the second gas supply unit 73, as described above. Therefore, the window 12 side of the second gas supply unit 73 is configured so that the laser light 301 can pass therethrough. Tin fine particles and tin ions are generated when the target substance forming the droplet DL is turned into plasma in the plasma generation region AR. The etching gas supplied from the first gas supply unit 72 and the second gas supply unit 73 contains hydrogen that reacts with the tin fine particles and the tin ions. Specifically, when the tin fine particles and the tin ions react with hydrogen, tin becomes stannane (SnH4) gas at room temperature. Here, a supply gas amount adjustment unit (not shown) may be arranged at a pipe between the gas supply tank 74 and at least one of the first gas supply unit 72 and the second gas supply unit 73. Further, a pair of exhaust portions 81 are arranged at the chamber 10. The exhaust portions 81 are configured to exhaust residual gas in the chamber 10. As shown in FIG. 3, exhaust ports 81E of the respective exhaust portions 81 are formed, for example, at positions facing each other on the wall of the chamber 10. The residual gas includes fine particles and charged particles generated by turning of the target substance into plasma, products generated through the reaction of the fine particles and the charged particles with the etching gas, and unreacted etching gas. Some of the charged particles are neutralized in the chamber 10, and the residual gas contains the neutralized charged particles as well. The exhaust portions 81 from which the residual gas is exhausted are connected to exhaust pipes 81P. The exhaust pipes 81P are connected to an exhaust pump 85, the residual gas exhausted from the exhaust ports 81E flows into the exhaust pump 85 through the exhaust pipes 81P, and the residual gas is subjected to predetermined exhaust treatment at the exhaust pump 85. Further, the EUV light generation apparatus 100 includes a magnetic field generation unit 65. The magnetic field generation unit 65 is configured to generate a magnetic field ML for the charged particles generated in the plasma generation region AR to converge to the exhaust port 81E. The magnetic field generation unit 65 may, for example, be configured by a pair of electromagnets 65M arranged so as to sandwich the wall of the chamber 10 facing each other. The pair of electromagnets 65M are arranged such that the plasma generation region AR is positioned midway between the respective electromagnets 65M. The direction of the current flowing through the superconducting coil of one electromagnet 65M is the same as the direction of the current flowing through the superconducting coil of the other electromagnet 65M. When such currents are applied to a pair of the superconducting coils, the magnetic field ML is generated in which the magnetic flux density is highest in the vicinity of the respective electromagnets 65M, and the magnetic flux density becomes lower toward the plasma generation region AR. A magnetic field axis MA of the magnetic field ML preferably crosses a reflective light path of the EUV light concentrating mirror 50 and preferably passes through the plasma generation region AR. The magnetic field ML is called a mirror magnetic field in some cases. Here, the magnetic field generation unit 65 may generate a magnetic field for the charged particles to converge from one electromagnet 65M side to the other electromagnet 65M side through the plasma generation region AR. Further, the magnetic field generation unit 65 is configured by the pair of electromagnets 65M but may be configured by a pair of permanent magnets. Further, the electromagnets 65M or the permanent magnets that are magnets for generating the magnetic field may be arranged inside the chamber 10. Further, as described above, in the comparative example, the exhaust ports 81E are provided at positions facing each other on the wall of the chamber 10, and the charged particles are exhausted from the exhaust ports 81E. Therefore, in the example shown in FIG. 3, the pair of exhaust ports 81E are provided at positions facing each other along the magnetic field axis MA and through which the magnetic field axis MA passes in the chamber 10. At least one of the exhaust portions 81 may be provided with a trap mechanism such as a heater for trapping fine particles. 3.2 Operation Next, operation of the EUV light generation apparatus 100 of the comparative example will be described. In the EUV light generation apparatus 100, for example, at the time of new installation, maintenance, or the like, atmospheric air in the chamber 10 is exhausted. At this time, purging and exhausting inside the chamber 10 may be repeated for exhausting atmospheric components. For example, inert gas such as nitrogen (N2) or argon (Ar) is preferably used for the purge gas. Thereafter, when the pressure in the chamber 10 becomes equal to or lower than a predetermined pressure, the control unit 20 starts introducing the etching gas from the first gas supply unit 72 and the second gas supply unit 73 into the chamber 10. At this time, the control unit 20 may control the supply gas amount adjustment unit (not shown) and the exhaust pump 85 so that the pressure at the internal space of the chamber 10 is maintained at the predetermined pressure. Thereafter, the control unit 20 waits until a predetermined time elapses from the start of introduction of the etching gas. Further, the control unit 20 drives the electromagnets 65M of the magnetic field generation unit 65 to generate the magnetic field ML. Further, the control unit 20 causes the gas in the internal space of the chamber 10 to be exhausted from the exhaust portion 81 to the exhaust pump 85, and keeps the pressure at the internal space of the chamber 10 substantially constant based on the signal of the pressure at the internal space of the chamber 10 measured by the pressure sensor 26. At this time, the pressure at the internal space of the chamber 10 is, for example, in the range of 10 Pa to 160 Pa. In order to heat and maintain the target substance in the tank 41 at a predetermined temperature equal to or higher than the melting point, the control unit 20 causes the heater power source 45 to apply current to the heater 44 to increase temperature of the heater 44. At this time, based on the output from a temperature sensor (not shown), an amount of the current applied from the heater power source 45 to the heater 44 is adjusted to control the temperature of the target substance to a predetermined temperature. The predetermined temperature is, for example, in the range of 250° C. to 290° C. when tin is used as the target substance. Further, the control unit 20 causes the pressure adjuster 43 to control the pressure in the tank 41 so that the melted target substance is output through a nozzle hole of the nozzle 42 at a predetermined velocity. The target substance discharged from the hole of the nozzle 42 may be in the form of jet. At this time, the control unit 20 applies voltage having a predetermined waveform to the piezoelectric element 46 through the piezoelectric power source 47 to generate the droplet DL. Vibration of the piezoelectric element 46 can propagate via the nozzle 42 to the jet of the target substance to be output through the hole of the nozzle 42. The jet of the target substance is divided at a predetermined cycle by the vibration, and a liquid droplet DL is generated from the target substance. Further, the control unit 20 outputs a light emission trigger to the laser device 30. When the light emission trigger is input, the laser device 30 emits the pulse laser light 301. The emitted laser light 301 is incident on the laser light concentrating optical system 13 through the laser light delivery optical system 35 and the window 12. At this time, the control unit 20 controls the laser light manipulator 13C of the laser light concentrating optical system 13 such that the laser light 301 is concentrated in the plasma generation region AR. The control unit 20 causes the laser device 30 to emit the laser light 301 based on the signal from the target sensor 27 so that the droplet DL is irradiated with the laser light 301. Thus, the droplet DL is irradiated in the plasma generation region AR with the laser light 301 converged by the laser light concentrating mirror 13A. Light including EUV light is emitted from the plasma generated through the irradiation. Among the light including the EUV light generated in the plasma generation region AR, the EUV light 101 is concentrated at the intermediate focal point IF by the EUV light concentrating mirror 50, and then is incident on the exposure apparatus 200 from the connection portion 19. Therefore, it can be understood that the connection portion 19 is the emission port of the EUV light in the EUV light generation apparatus 100. Here, when the target substance is turned into plasma, charged particles are generated as described above. The charged particles move on a trajectory rotating in a plane perpendicular to the magnetic field line by receiving a Lorentz force from the magnetic field ML. When the charged particles moving in this manner have velocity components in the direction to the exhaust port 81E, the charged particles are directed toward the wall of the chamber 10 while converging in a spiral trajectory along the magnetic field lines. Accordingly, most of the charged particles are guided to the exhaust port 81E provided on the wall of the chamber 10 near the converging portion of the magnetic field ML, and flow into the exhaust port 81E. The charged particles reaching the inside of the exhaust pipe 81P from the exhaust port 81E flow into the exhaust pump 85 along with airflow of the exhaust gas. Further, when the target substance is turned into plasma, electrically neutral fine particles are generated in addition to the charged particles described above. Not being affected by the magnetic field ML generated by the magnetic field generation unit 65, the fine particles are diffused into the chamber 10. Some of the fine particles diffused into the chamber 10 adhere to a reflection surface of the EUV light concentrating mirror 50. The fine particles adhered to the reflection surface react with the etching gas supplied from the first gas supply unit 72 and the second gas supply unit 73, and become predetermined products through the reaction. As described above, when tin is used as the target substance and gas containing hydrogen is used as the etching gas, the product is stannane gas at room temperature. Most of the products generated by the reaction with the etching gas flow into the exhaust pipe 81P from the exhaust port 81E of the exhaust portion 81 along with the flow of the unreacted etching gas. Further, at least some of the charged particles not converged to the exhaust port 81E and the fine particles not adhered to the reflection surface of the EUV light concentrating mirror 50 may react with some of the unreacted etching gas flowing in the chamber 10. Most of the products generated by the reaction flow into the exhaust pipe 81P from the exhaust port 81E along with the flow of the unreacted etching gas. At least some of the unreacted etching gas flows into the exhaust pipe 81P from the exhaust port 81E. The unreacted etching gas, fine particles, charged particles, neutralized charged particles, the above-mentioned products, and the like, which have flowed into the exhaust port 81E of the exhaust portion 81, flow from the exhaust pipe 81P into the exhaust pump 85 as exhaust gas and are subjected to predetermined exhaust treatment such as detoxification. 3.3 Problem Most of the charged particles and fine particles generated by turning of the droplet DL into plasma are discharged from the exhaust port 81E as described above. However, some of the charged particles and fine particles in the chamber 10 remain in the chamber 10 without being discharged from the exhaust port 81E in some cases. When the charged particles remain in the chamber 10, the charged particles tend to be neutralized. Further, the charged particles remaining in the chamber 10 may collide with the fine particles in the chamber 10 to generate new fine particles. When the fine particles remain in the chamber 10, there is a strong tendency for the fine particles to deposit on the reflection surface 55 of the EUV light concentrating mirror 50. Consequently, there may arise a problem that the reflectance of the EUV light concentrating mirror 50 is reduced. When the reflectance of the EUV light concentrating mirror 50 is reduced, there is a fear that output of the EUV light from the EUV light generation apparatus 100 is reduced. Therefore, in the following embodiments, an extreme ultraviolet light generation apparatus capable of suppressing decrease in reflectance of EUV light is exemplified. 4. Description of Extreme Ultraviolet Light Generation Apparatus of First Embodiment Next, the configuration of an extreme ultraviolet light generation apparatus of a first embodiment will be described. Any component same as that described above is denoted by an identical reference sign, and duplicate description thereof is omitted unless specific description is needed. 4.1 Configuration FIG. 4 is a schematic view showing a schematic configuration of a part including the chamber 10 of the EUV light generation apparatus 100 in the present embodiment. In FIG. 4, similarly to FIG. 3, the laser light concentrating optical system 13 is omitted. The chamber 10 of the present embodiment is mainly different from the chamber 10 of the comparative example in that a second exhaust portion 83 is provided. In the following description, the exhaust portion 81 described in the comparative example is referred to as a first exhaust portion 81, the exhaust port 81E is referred to as a first exhaust port 81E, and the exhaust pipe 81P is referred to as a first exhaust pipe 81P. In the present embodiment, since the first exhaust portion 81 and the second exhaust portion 83 are provided, the residual gas in the chamber 10 is exhausted as the first exhaust gas from the first exhaust portion 81 and exhausted as the second exhaust gas from the second exhaust portion 83. As shown in FIG. 4, the second exhaust portion 83 includes a second exhaust port 83E. The second exhaust port 83E is arranged at a position opposite to the EUV light concentrating mirror 50 in the chamber 10 with respect to a plane passing through the first exhaust port 81E and being perpendicular to the optical axis of the EUV light concentrating mirror 50. The pair of first exhaust ports 81E are arranged at positions through which the magnetic field axis MA passes in the chamber 10 as described above. Therefore, the second exhaust port 83E is arranged on a side close to the connection portion 19 being the emission port of the EUV light not on a side close to the plasma generation region AR, provided that the magnetic field axis MA crosses the reflected light path of the EUV light concentrating mirror 50 as passing through the plasma generation region AR. The second exhaust port 83E is connected to the second exhaust pipe 83P. The second exhaust pipe 83P is connected to a gas exhaust amount adjustment unit 86. Further, in the present embodiment, the first exhaust pipe 81P is also connected to the gas exhaust amount adjustment unit 86, similarly to the second exhaust pipe 83P. The gas exhaust amount adjustment unit 86 is connected to an exhaust pipe 87. The exhaust pipe 87 is connected to the exhaust pump 85. The gas exhaust amount adjustment unit 86 adjusts the ratio between the first exhaust gas exhausted from the first exhaust pipe 81P to the exhaust pipe 87 and the second exhaust gas exhausted from the second exhaust pipe 83P to the exhaust pipe 87. As described above, the first exhaust pipe 81P is connected to the first exhaust port 81E, and the second exhaust pipe 83P is connected to the second exhaust port 83E. Therefore, the gas exhaust amount adjustment unit 86 adjusts the ratio between the first exhaust gas exhausted from the first exhaust port 81E and the second exhaust gas exhausted from the second exhaust port 83E. In the present embodiment, when a path through which the first exhaust gas flows and through which the flow of the second exhaust gas is suppressed is referred to as a first exhaust path, a path from the first exhaust port 81E to the gas exhaust amount adjustment unit 86 corresponds to the first exhaust path. Further, when a path through which the second exhaust gas flows and through which the flow of the first exhaust gas is suppressed is referred to as a second exhaust path, a path from the second exhaust port 83E to the gas exhaust amount adjustment unit 86 corresponds to the second exhaust path. The exhaust pump 85 is a pump for exhausting gas from the exhaust pipe 87. Accordingly, the exhaust pump 85 exhausts a part of the residual gas in the chamber 10 from the first exhaust port 81E to the first exhaust pipe 81P as the first exhaust gas, and exhausts another part of the residual gas from the second exhaust port 83E to the second exhaust pipe 83P as the second exhaust gas, through the exhaust pipe 87 and the gas exhaust amount adjustment unit 86. Further, in the present embodiment, the supply gas amount adjustment unit 76 is arranged in the pipe between the gas supply tank 74 and the first gas supply unit 72 and in the pipe between the gas supply tank 74 and the second gas supply unit 73. The supply gas amount adjustment unit 76 adjusts the flow rate of the etching gas flowing from the gas supply tank 74 to the first gas supply unit 72 and the flow rate of the etching gas flowing from the gas supply tank 74 to the second gas supply unit 73. In addition, in the present embodiment, the chamber 10 is provided with an EUV light measurement unit 28. The EUV light measurement unit 28 detects the intensity of part of the EUV light radiated from the plasma generation region AR among the light generated in the plasma generation region AR, and outputs a signal corresponding to the intensity. Further, in the present embodiment, the EUV light measurement unit 28, the gas exhaust amount adjustment unit 86, and the supply gas amount adjustment unit 76 are connected to the control unit 20. In the present embodiment, the control unit 20 controls the gas exhaust amount adjustment unit 86 and the supply gas amount adjustment unit 76 in addition to the operation in the comparative example. A memory 25 is connected to the control unit 20. The memory 25 stores a table in which the amount of the etching gas supplied from the first gas supply unit 72 and the amount of the etching gas supplied from the second gas supply unit 73 are associated with each other. The amount of the etching gas is determined based on the intensity of the laser light 301 emitted from the laser device 30. Therefore, the memory 25 further stores a table in which the intensity of the laser light 301 and the amount of the etching gas supplied into the chamber 10 are associated with each other. In the present embodiment, the higher the intensity of the laser light 301, the larger the amount of the etching gas supplied into the chamber 10. In the present embodiment, the amount of the etching gas supplied from the first gas supply unit 72 is constant regardless of the intensity of the laser light 301. Therefore, only the amount of the etching gas supplied from the second gas supply unit 73 in accordance with the intensity of the laser light 301 may be stored in the table stored in the memory 25. Here, unlike the present embodiment, the amount of the etching gas supplied from the first gas supply unit 72 may vary in accordance with the intensity of the laser light 301. The memory 25 stores the ratio between the amount of the first exhaust gas exhausted from the first exhaust portion 81 and the amount of the second exhaust gas exhausted from the second exhaust portion 83. The ratio is determined based on the amount of the etching gas supplied into the chamber 10. Accordingly, the memory 25 stores a table in which the amount of the etching gas and the ratio are associated with each other. For example, the ratio of the amount of the first exhaust gas to the amount of the second exhaust gas is determined to be 70%:30% to 30%:70%. 4.2 Operation Next, operation of the EUV light generation apparatus 100 of the present embodiment will be described. FIG. 5 is a flowchart showing operation of the control unit 20 from an initial state until the EUV light generation apparatus 100 emits EUV light. (Step SP11) Similarly to the comparative example, for example, at the time of new introduction, maintenance, or the like, the atmosphere in the chamber 10 is exhausted and the pressure in the chamber 10 is reduced in step SP11. (Step SP12) When the pressure in the chamber 10 becomes equal to or lower than a predetermined pressure defined as the initial state, in this step, the control unit 20 starts supplying the etching gas from the first gas supply unit 72 and the second gas supply unit 73 into the chamber 10. At this stage, the laser device 30 does not emit the laser light 301. However, in the present embodiment, the control unit 20 controls the supply gas amount adjustment unit 76 with reference to the memory 25 so that the amount of the etching gas becomes that when the laser light 301 is emitted from the laser device 30 at predetermined intensity in step SP17 described later. At this time, the control unit 20 controls the supply gas amount adjustment unit 76 with reference to the memory 25 so that the ratio between the amount of the etching gas supplied from the first gas supply unit 72 and the amount of the etching gas supplied from the second gas supply unit 73 becomes a predetermined ratio. The supply gas amount adjustment unit 76 adjusts the amount of the etching gas supplied from the first gas supply unit 72 and the amount of the etching gas supplied from the second gas supply unit 73, based on the control signal from the control unit 20. (Step SP13) In this step, the control unit 20 controls the gas exhaust amount adjustment unit 86 so that the same amount of gas as the amount of the etching gas supplied into the chamber 10 is exhausted. At this time, the control unit 20 refers to the table stored in the memory 25 and controls the gas exhaust amount adjustment unit 86 so that the ratio between the amount of the first exhaust gas and the amount of the second exhaust gas becomes a predetermined ratio. As described above, the ratio between the amount of the first exhaust gas and the amount of the second exhaust gas is determined based on the amount of the etching gas supplied into the chamber 10. Thus, the predetermined ratio at this time corresponds to the ratio between the amounts of the etching gas supplied into the chamber 10 when the laser light 301 is emitted from the laser device 30 at the predetermined intensity in step SP17 described later. (Step SP14) In this step, based on the signal from the pressure sensor 26, the control unit 20 proceeds to step SP15 when the pressure in the chamber 10 is within a predetermined pressure range, and returns to step SP12 when the pressure in the chamber 10 is not within the predetermined pressure range. At this time, the pressure at the internal space of the chamber 10 is, for example, in the range of 10 Pa to 160 Pa. (Step SP15) In this step, similarly to the comparative example, the control unit 20 drives the electromagnets 65M of the magnetic field generation unit 65 to generate the magnetic field ML. (Step SP16) In this step, similarly to the comparative example, the control unit 20 controls the pressure in the tank 41 by the pressure adjuster 43 so that the molten target substance is output from the nozzle hole of the nozzle 42 at a predetermined velocity. (Step SP17) In this step, similarly to the comparative example, the control unit 20 outputs the light emission trigger to the laser device 30. When the emission trigger is input, the laser device 30 emits the pulse laser light 301 having predetermined intensity. Similarly to the comparative example, the emitted laser light 301 is concentrated in the plasma generation region AR and is radiated to the droplet DL 301. At this time, the amount of the etching gas supplied from the first gas supply unit 72 and the amount of the etching gas supplied from the second gas supply unit 73 remain the same as the amounts of the etching gas supplied in step SP12. (Step SP18) In this step, light including EUV light is radiated from the plasma generated through the irradiation of the droplet DL with the laser light 301 in step SP17. Similarly to the comparative example, the EUV light 101 among the light including the EUV light generated in the plasma generation region AR enters the exposure apparatus 200 from the connection portion 19. Next, description will be provided on operation of the EUV light generation apparatus 100 in a state that the EUV light 101 is emitted from the EUV light generation apparatus 100. FIG. 6 is a flowchart showing operation of the control unit 20 in a state that the EUV light 101 is emitted from the EUV light generation apparatus 100. (Step SP21) The state in which the EUV light 101 is emitted from the EUV light generation apparatus 100 in step SP18 of FIG. 5 is a state of start shown in FIG. 6. In this step, the intensity of the EUV light 101 emitted from the EUV light generation apparatus 100 is measured. Specifically, the EUV light measurement unit 28 detects the intensity of part of the EUV light radiated from the plasma generation region AR among the light generated in the plasma generation region AR, and outputs a signal corresponding to the intensity. The output signal is input to the control unit 20. The intensity of the EUV light measured by the EUV light measurement unit 28 differs from the intensity of the EUV light 101 output from the connection portion 19 of EUV light generation apparatus 100. However, the intensity of the EUV light radiated from the plasma generation region AR and detected by the EUV light measurement unit 28 is correlated with the intensity of the EUV light 101 output from the connection portion 19 of the EUV light generation apparatus 100. Therefore, in the present embodiment, using the intensity of the EUV light radiated from the plasma generation region AR and detected by the EUV light measurement unit 28, the intensity of the EUV light 101 output from the connection portion 19 of the EUV light generation apparatus 100 is measured. (Step SP22) In this step, the control unit 20 determines whether or not the intensity of the EUV light 101 output from the connection portion 19 is within a predetermined range using the signal input from the EUV light measurement unit 28. The control unit 20 may perform the above determination by calculating the intensity of the EUV light 101 output from the connection portion 19 of the EUV light generation apparatus 100 using the signal input from the EUV light measurement unit 28. Alternatively, the control unit 20 determines whether or not the intensity of the EUV light detected by the EUV light measurement unit 28 is within a predetermined range, and whether or not the intensity of the EUV light 101 output from the connection portion 19 is within the predetermined range based on the above determination. When the intensity of the EUV light 101 is within the predetermined range, there is no need to change the EUV light 101 output from the connection portion 19 of the EUV light generation apparatus 100. Therefore, there is no need to change the laser light 301 emitted from the laser device 30. In this case, the control unit 20 returns to step SP21, and receives a signal output from the EUV light measurement unit 28. On the other hand, when the control unit 20 determines that the intensity of the EUV light 101 is not within the predetermined range, the control unit 20 proceeds to step SP23. (Step SP23) In this step, the control unit 20 determines whether or not to stop emitting the laser light 301 from the laser device 30. In step SP22, the control unit 20 determines whether or not the difference of the intensity of the EUV light 101 output from the connection portion 19 from the above predetermined range is equal to or greater than a certain intensity. When the difference of the intensity of the EUV light 101 from the predetermined range is equal to or greater than the certain intensity, that is, when the intensity of the EUV light 101 output from the connection portion 19 differs greatly from the assumed range, there is a possibility that some failure has occurred in the EUV light generation apparatus 100. In this case, the control unit 20 controls the laser device 30 to stop emitting the laser light 301 therefrom and proceeds to step SP28. In step SP28, the EUV light generation apparatus 100 is stopped. On the other hand, when the control unit 20 determines that the difference of the intensity of the EUV light 101 from the predetermined range is less than the certain intensity, the control unit 20 proceeds to step SP24. (Step SP24) In this step, the control unit 20 calculates the intensity of the laser light 301 emitted from laser device 30 required for the intensity of the EUV light 101 to be within the predetermined range. For example, as described above, when the target substance is turned into plasma, particles such as charged particles and electrically neutral fine particles are generated, and some of these particles may adhere to the reflection surface of the EUV light concentrating mirror 50. Then, there may be a case that the intensity of the EUV light 101 output from the connection portion 19 falls below the predetermined range while the reflectance of the EUV light concentrating mirror 50 is decreased. When the intensity of the EUV light 101 is lower than the predetermined range, the control unit 20 performs the calculation to increase the intensity of the laser light 301. On the other hand, when the intensity of the EUV light 101 is higher than the predetermined range, the control unit 20 performs the calculation to decrease the intensity of the laser light 301. Thus, the intensity of the laser light 301 to set the intensity of the EUV light 101 within the predetermined range is obtained. (Step SP25) In this step, based on the intensity of the laser light 301 calculated in step SP24, the amount of the etching gas to be supplied is determined. In the case that the intensity of the laser light 301 is increased, charged particles, electrically neutral fine particles, and the like tend to be emitted more when the laser light 301 is radiated to the target substance. Therefore, when the intensity of the laser light 301 is increased, the amount of the etching gas supplied from the second gas supply unit 73 is increased, and when the intensity of the laser light 301 is decreased, the amount of the etching gas supplied from the second gas supply unit 73 is decreased. As described above, in the present embodiment, the amount of the gas supplied from the first gas supply unit 72 is constant regardless of the intensity of the laser light 301. Accordingly, in this step, the amount of the etching gas supplied from the second gas supply unit 73 is determined based on the table stored in the memory 25. Therefore, in this step, the ratio between the amount of the gas supplied from the first gas supply unit 72 and the amount of the etching gas supplied from the second gas supply unit 73 is consequently determined. (Step SP26) In this step, based on the amount of the etching gas determined in step SP25, the ratio between the amount of the first exhaust gas exhausted from the first exhaust portion 81 and the amount of the second exhaust gas exhausted from the second exhaust portion 83 is determined. The control unit 20 determines the ratio by reading out from the table in which the ratio is associated with the amount of the etching gas in the memory 25. When the memory 25 does not store such a table, the control unit 20 may calculate the ratio from the amount of the etching gas based on a predetermined algorithm. (Step SP27) In this step, the control unit 20 controls the laser device 30 so that the laser light 301 having the intensity calculated in step SP24 is emitted from the laser device 30. Thus, the laser device 30 emits the pulse laser light 301 having the intensity calculated in step SP24. The laser light 301 having the new intensity is concentrated in the plasma generation region AR and radiated to the target substance. Accordingly, the EUV light 101 having the intensity within the predetermined range is output from the connection portion 19. For example, when the reflectance of the EUV light concentrating mirror 50 is decreased, the new intensity of the laser light 301 will be greater than the previous intensity of the laser light 301, and the total amount of the EUV light radiated from the plasma generation region AR is increased. Therefore, even when the reflectance of the EUV light concentrating mirror 50 is decreased, the EUV light 101 having the intensity within the predetermined range can be output from the connection portion 19. Further, in this step, the control unit 20 controls the supply gas amount adjustment unit 76 so that the amount of the etching gas to be supplied from the second gas supply unit 73 determined in step SP25 is supplied from the second gas supply unit 73. The supply gas amount adjustment unit 76 supplies the etching gas from the second gas supply unit 73 based on the control of the control unit 20. For example, when the new intensity of the laser light 301 is greater than the previous intensity of the laser light 301, the amount of the etching gas supplied from the second gas supply unit 73 is set greater than the amount of the etching gas previously supplied from the second gas supply unit 73. Further, in this step, the control unit 20 controls the gas exhaust amount adjustment unit 86 so that the ratio between the amount of the first exhaust gas exhausted from the first exhaust portion 81 and the amount of the second exhaust gas exhausted from the second exhaust portion 83 becomes the ratio determined in step SP26. Accordingly, the first exhaust gas and the second exhaust gas are exhausted from the first exhaust portion 81 and the second exhaust portion 83 at the ratio determined in step SP26. The first exhaust gas and the second exhaust gas exhausted as described above are exhausted to the exhaust pump 85 through the exhaust pipe 87. Here, the control unit 20 controls the gas exhaust amount adjustment unit 86 so that the total amount of the first exhaust gas and the second exhaust gas is equal to the total amount of the etching gas supplied from the first gas supply unit 72 and the second gas supply unit 73, as described above. Accordingly, the pressure in the chamber 10 is maintained substantially constant. Next, returning to step SP21, the intensity of the EUV light measured by the EUV light measurement unit 28 is input to the control unit 20. 4.3 Effect In the present embodiment, the first exhaust port 81E is arranged at a position through which the magnetic field axis MA passes in the chamber 10. Further, the second exhaust port 83E is arranged at a position opposite to the EUV light concentrating mirror 50 in the chamber 10 with respect to a plane passing through the first exhaust port 81E and being perpendicular to the optical axis of the EUV light concentrating mirror 50. Further, the EUV light generation apparatus 100 includes the gas exhaust amount adjustment unit 86 that adjusts the ratio between the exhaust amount of the first exhaust gas exhausted from the first exhaust port 81E and the exhaust amount of the second exhaust gas exhausted from the second exhaust port 83E. Therefore, compared with the case where the first exhaust port 81E is arranged in the chamber 10 and the second exhaust port 83E is not arranged as in the EUV light generation apparatus 100 of the comparative example, it is possible to make it easier for the residual gas in the chamber 10 to flow toward the side opposite to the EUV light concentrating mirror 50. Further, the gas exhaust amount adjustment unit 86 adjusts the ratio between the exhaust amount of the first exhaust gas exhausted from the first exhaust port 81E and the exhaust amount of the second exhaust gas exhausted from the second exhaust port 83E. Accordingly, it is possible to adjust the amount of the residual gas in the chamber 10 flowing toward the side opposite to the EUV light concentrating mirror 50. Thus, as compared with the EUV light generation apparatus 100 of the comparative example, it is possible to suppress contamination of the EUV light concentrating mirror 50 with fine particles generated at the time of EUV light generation, and it is possible to suppress decrease in reflectance of EUV light. Further, in the EUV light generation apparatus 100 of the present embodiment, the pair of first exhaust ports 81E are arranged at the chamber 10 at positions facing each other along the magnetic field axis MA. Therefore, charged particles flowing toward the wall of the chamber 10 can be effectively exhausted from the first exhaust port 81E while converging in a spiral trajectory along the magnetic field lines. Although the first exhaust ports 81E are arranged as a pair in the present embodiment, for example, one first exhaust port 81E may be arranged in the chamber 10. Further, in the EUV light generation apparatus 100, the control unit 20 controls the gas exhaust amount adjustment unit 86 to adjust the ratio between the exhaust amount of the first exhaust gas and the exhaust amount of the second exhaust gas based on the amount of the etching gas supplied into the chamber 10. When the amount of the etching gas supplied into the chamber 10 varies, the flow of the gas in the chamber 10 is changed. Owing to that the control unit 20 controls the gas exhaust amount adjustment unit 86 to adjust the ratio based on the amount of the etching gas supplied into the chamber 10, the amount of the residual gas in the chamber 10 flowing toward the side opposite to the EUV light concentrating mirror 50 can be more appropriately adjusted. Further, in the EUV light generation apparatus 100 of the present embodiment, the control unit 20 controls the supply gas amount adjustment unit 76 to adjust the amount of the etching gas supplied into the chamber 10 based on the intensity of the laser light 301. When the intensity of the laser light 301 is high, fine particles are more easily scattered from the target substance irradiated with the laser light 301 than when the intensity of the laser light 301 is low. Therefore, by adjusting the amount of the etching gas supplied into the chamber 10 based on the intensity of the laser light 301, the residual gas containing the scattered fine particles can be more effectively exhausted from the chamber 10. Further, in the EUV light generation apparatus 100 of the present embodiment, the control unit 20 controls the supply gas amount adjustment unit 76 to adjust the amount of the etching gas supplied from the second gas supply unit 73. The second gas supply unit 73 supplies the etching gas in a direction away from the EUV light concentrating mirror 50. Accordingly, by adjusting the amount of the etching gas supplied from the second gas supply unit 73, the amount of the fine particles scattered from the target substance toward the EUV light concentrating mirror 50 can be appropriately reduced. Here, the second gas supply unit 73 may supply the etching gas in a direction other than the direction away from the EUV light concentrating mirror 50. Although the etching gas is described as an example in the above description, the gas supplied from the first gas supply unit 72 and the second gas supply unit 73 is not limited to the etching gas, and may be, for example, inert gas, mixed gas of etching gas and inert gas, or the like. 5. Description of Extreme Ultraviolet Light Generation Apparatus of Second Embodiment Next, an extreme ultraviolet light generation apparatus of a second embodiment will be described. Any component same as that described above is denoted by an identical reference sign, and duplicate description thereof is omitted unless specific description is needed. 5.1 Configuration FIG. 7 is a schematic view showing a schematic configuration of a part including the chamber 10 of the EUV light generation apparatus 100 in the present embodiment in the same manner as FIG. 4. As shown in FIG. 7, in the EUV light generation apparatus 100 of the present embodiment, a first exhaust pump 85A that exhausts the first exhaust gas from the first exhaust port 81E is connected to the first exhaust pipe 81P and a first valve 81V is arranged in the first exhaust pipe 81P. Further, in the EUV light generation apparatus 100 of the present embodiment, a second exhaust pump 85B that exhausts the second exhaust gas from the second exhaust port 83E is connected to the second exhaust pipe 83P and a second valve 83V is arranged in the second exhaust pipe 83P. Here, as described in the first embodiment, a path through which the first exhaust gas flows and through which the flow of the second exhaust gas is suppressed is referred to as the first exhaust path, and a path through which the second exhaust gas flows and through which the flow of the first exhaust gas is suppressed is referred to as the second exhaust path. In the present embodiment, the path from the first exhaust port 81E to the first exhaust pump 85A corresponds to the first exhaust path, and the path from the second exhaust port 83E to the second exhaust pump 85B corresponds to the second exhaust path. Accordingly, the first valve 81V is arranged in the first exhaust path and the second valve 83V is arranged in the second exhaust path. Further, in the present embodiment, the gas exhaust amount adjustment unit 86 includes the first valve 81V and the second valve 83V. The opening of the first valve 81V and the opening of the second valve 83V are controlled by the control unit 20. 5.2 Operation The EUV light generation apparatus 100 of the present embodiment operates in the same manner as the EUV light generation apparatus 100 of the first embodiment described with reference to FIGS. 5 and 6. However, in the present embodiment, the ratio between the first exhaust gas and the second exhaust gas in step SP13 and step SP27 is adjusted by the first valve 81V and the second valve 83V, each opening of which is controlled by the control unit 20. 5.3 Effect In the EUV light generation apparatus 100 of the present embodiment, due to the opening of the first valve 81V and the opening of the second valve 83V, it is possible to adjust the ratio between the amount of the first exhaust gas exhausted from the first exhaust port 81E and the amount of the second exhaust gas exhausted from the second exhaust port 83E. Thus, the ratio can be finely adjusted. Further, the first exhaust pump 85A is connected to the first exhaust pipe 81P and the second exhaust pump 85B is connected to the second exhaust pipe 83P. That is, the first exhaust path and the second exhaust path are independently connected to the exhaust pumps, respectively. Therefore, when the amount of the first exhaust gas is adjusted, the influence thereof on the amount of the second exhaust gas can be suppressed, and when the amount of the second exhaust gas is adjusted, the influence thereof on the amount of the first exhaust gas can be suppressed. 6. Description of Extreme Ultraviolet Light Generation Apparatus of Third Embodiment Next, an extreme ultraviolet light generation apparatus of a third embodiment will be described. Any component same as that described above is denoted by an identical reference sign, and duplicate description thereof is omitted unless specific description is needed. 6.1 Configuration FIG. 8 is a schematic view showing a schematic configuration of a part including the chamber 10 of the EUV light generation apparatus 100 of the present embodiment in the same manner as FIG. 4. As shown in FIG. 8, in the EUV light generation apparatus 100 of the present embodiment, the first exhaust pipe 81P connected to the first exhaust port 81E and the second exhaust pipe 83P connected to the second exhaust port 83E are connected respectively to one end of the exhaust pipe 87, and the other end of the exhaust pipe 87 is connected to the exhaust pump 85. Here, as described in the first embodiment, a path through which the first exhaust gas flows and through which the flow of the second exhaust gas is suppressed is referred to as the first exhaust path, and a path through which the second exhaust gas flows and through which the flow of the first exhaust gas is suppressed is referred to as the second exhaust path. In this case, in the present embodiment, the path from the first exhaust port 81E to the front of the exhaust pipe 87 corresponds to the first exhaust path, and the path from the second exhaust port 83E to the front of the exhaust pipe 87 corresponds to the second exhaust path. Further, similarly to the second embodiment, the first valve 81V and the second valve 83V are arranged in the first exhaust pipe 81P and the second exhaust pipe 83P, respectively. Accordingly, in the present embodiment as well, the first valve 81V is arranged in the first exhaust path and the second valve 83V is arranged in the second exhaust path. Further, in the present embodiment, similarly to the second embodiment, the gas exhaust amount adjustment unit 86 includes the first valve 81V and the second valve 83V. The opening of the first valve 81V and the opening of the second valve 83V are controlled by the control unit 20. 6.2 Operation Similarly to the second embodiment, the EUV light generation apparatus 100 of the present embodiment operates similarly to the EUV light generation apparatus 100 of the first embodiment described with reference to FIGS. 5 and 6. However, in the present embodiment, the first exhaust gas from the first exhaust port 81E and the second exhaust gas from the second exhaust port 83E are exhausted by the exhaust pump 85. 6.3 Effect Similarly to the second embodiment, in the EUV light generation apparatus 100 of the present embodiment, due to the opening of the first valve 81V and the opening of the second valve 83V, it is possible to adjust the ratio between the amount of the first exhaust gas exhausted from the first exhaust port 81E and the amount of the second exhaust gas exhausted from the second exhaust port 83E. Thus, the ratio can be finely adjusted. Further, the first exhaust gas and the second exhaust gas are exhausted commonly by the exhaust pump 85. Therefore, compared with the EUV light generation apparatus 100 of the second embodiment, the configuration can be simplified. 7. Description of Extreme Ultraviolet Light Generation Apparatus of Fourth Embodiment Next, an extreme ultraviolet light generation apparatus of a fourth embodiment will be described. Any component same as that described above is denoted by an identical reference sign, and duplicate description thereof is omitted unless specific description is needed. 7.1 Configuration FIG. 9 is a schematic view showing a schematic configuration of a part including the chamber 10 of the EUV light generation apparatus 100 of the present embodiment in the same manner as FIG. 4. As shown in FIG. 9, in the EUV light generation apparatus 100 of the present embodiment, similarly to the EUV light generation apparatus 100 of the third embodiment, the first exhaust pipe 81P connected to the first exhaust port 81E and the second exhaust pipe 83P connected to the second exhaust port 83E are connected respectively to one end of the exhaust pipe 87. Here, as described in the first embodiment, a path through which the first exhaust gas flows and through which the flow of the second exhaust gas is suppressed is referred to as the first exhaust path, and a path through which the second exhaust gas flows and through which the flow of the first exhaust gas is suppressed is referred to as the second exhaust path. In the present embodiment, the path from the first exhaust port 81E to the front of the exhaust pipe 87 corresponds to the first exhaust path, and the path from the second exhaust port 83E to the front of the exhaust pipe 87 corresponds to the second exhaust path. The other end of the exhaust pipe 87 is connected to the exhaust pump 85. Further, similarly to the second embodiment and the third embodiment, the second valve 83V is arranged in the second exhaust pipe 83P. Accordingly, in the present embodiment, the second valve 83V is arranged in the second exhaust path. However, in the present embodiment, no valve is arranged in the first exhaust pipe 81P. Accordingly, no valve is arranged in the first exhaust path. In the present embodiment, the gas exhaust amount adjustment unit 86 includes the second valve 83V. The opening of the second valve 83V is controlled by the control unit 20. 7.2 Operation Similarly to the second embodiment, the EUV light generation apparatus 100 of the present embodiment operates similarly to the EUV light generation apparatus 100 of the first embodiment described with reference to FIGS. 5 and 6. However, in the present embodiment, in steps SP13 and SP27, the ratio between the first exhaust gas exhausted from the first exhaust port 81E and the second exhaust gas exhausted from the second exhaust port 83E is adjusted by the second valve 83V, the openings of which are controlled by the control unit 20. Since no valve is arranged in the first exhaust path, the amount of the first exhaust gas exhausted from the first exhaust port 81E is substantially constant. 7.3 Effect In the EUV light generation apparatus 100 of the present embodiment, the first exhaust gas and the second exhaust gas are exhausted commonly by the exhaust pump 85. Therefore, similarly to the EUV light generation apparatus 100 of the third embodiment, the configuration can be simplified. Further, in the present embodiment, although the second valve 83V is arranged in the second exhaust path, the EUV light generation apparatus 100 may have a simpler configuration since no valve is arranged in the first exhaust path. The description above is intended to be illustrative and the present disclosure is not limited thereto. Therefore, it would be obvious to those skilled in the art that various modifications to the embodiments of the present disclosure would be possible without departing from the spirit and the scope of the appended claims. Further, it would be also obvious to those skilled in the art that embodiments of the present disclosure would be appropriately combined. The terms used throughout the present specification and the appended claims should be interpreted as non-limiting terms unless clearly described. For example, terms such as “comprise”, “include”, “have”, and “contain” should not be interpreted to be exclusive of other structural elements. Further, indefinite articles “a/an” described in the present specification and the appended claims should be interpreted to mean “at least one” or “one or more.” Further, “at least one of A, B, and C” should be interpreted to mean any of A, B, C, A+B, A+C, B+C, and A+B+C as well as to include combinations of the any thereof and any other than A, B, and C.
043026800	abstract	There is provided a cover construction for a shielding container for the transportation and storage of fuel elements essentially consisting of an inner shielding cover and an outer transportation cover flanged on the cover, the inner shielding cover consisting of a fixed cover portion and a moveable cover portion provided at the rim of the cover with claws, whereby the claws are fitted into corresponding openings of the top of the container and the connection between the fixed cover portion and the moveable cover portion takes place through a screw carried into the moveable portion and a pin fitting into the fixed portion.
	claims	1. Apparatus for scanning a charged particle beam, comprising: scan elements spaced apart by a gap for passing a charged particle beam; a scan signal generator coupled to said scan elements for generating scan signals for scanning the charged particle beam in a scan pattern having a scan origin; and a position controller for positioning said scan elements based on at least one parameter of the charged particle beam. 2. Apparatus as defined in claim 1 wherein said scan elements comprise electrostatic scan plates for electrostatic deflection of the charged particle beam and wherein said scan signal generator comprises a scan voltage generator. claim 1 3. Apparatus as defined in claim 1 wherein said scan elements comprise magnetic polepieces and a magnet coil for energizing said magnetic polepieces, and wherein scan signal generator comprises a scan current generator for energizing said magnet coil. claim 1 4. Apparatus as defined in claim 1 wherein said position controller comprises means for positioning said scan elements to achieve a desired position of the scan origin for given parameter values of the charged particle beam. claim 1 5. Apparatus as defined in claim 1 wherein said position controller comprises means for positioning said scan elements to achieve a fixed position of the scan origin for different parameter values of the charged particle beam. claim 1 6. Apparatus as defined in claim 1 wherein said position controller comprises means for positioning said scan elements to achieve a desired position of the scan origin as the energy of the charged particle beam changes. claim 1 7. Apparatus as defined in claim 2 wherein said position controller comprises means for moving said scan plates upstream with respect to the charged particle beam as the spacing between the scan plates is increased. claim 2 8. Apparatus as defined in claim 2 wherein said position controller comprises means for moving said scan plates upstream along linear paths at angles with respect to an axis of the charged particle beam. claim 2 9. Apparatus as defined in claim 2 wherein said position controller comprises means for rotating said scan plates as the spacing between said scan plates is changed. claim 2 10. Apparatus as defined in claim 2 wherein said position controller comprises means for translating said scan plates along linear paths disposed at opposite angles with respect to an axis of the charged particle beam. claim 2 11. Apparatus as defined in claim 1 wherein said position controller comprises means for automatically positioning said scan elements based on the parameter of the charged particle beam. claim 1 12. Apparatus as defined in claim 2 wherein said position controller comprises means for moving said scan plates to two or more discrete positions based on the energy of the charged particle beam. claim 2 13. Apparatus as defined in claim 2 wherein said position controller comprises means for moving said scan plates along a continuous range of positions based on the energy of the charged particle beam. claim 2 14. Apparatus for scanning a charged particle beam, comprising: first scan elements spaced apart by a first gap for passing a charged particle beam; second scan elements spaced apart by a second gap for passing the charged particle beam; a scan signal generator coupled to said scan elements for generating scan signals for scanning the charged particle beam in a scan pattern having a scan origin; and a scan signal controller for controlling the scan signals supplied from said scan signal generator to said first scan elements and said second scan elements based on at least one parameter of the charged particle beam. 15. Apparatus as defined in claim 14 wherein said first scan elements and said second scan elements each comprise scan plates for electrostatic deflection of the charged particle beam and wherein said scan signal generator comprises a scan voltage generator. claim 14 16. Apparatus as defined in claim 14 wherein said first scan elements and said second scan elements each comprise magnetic polepieces and a magnet coil for energizing said magnetic polepieces, and wherein said scan signal generator comprises a scan current generator for energizing said magnet coil. claim 14 17. Apparatus as defined in claim 14 wherein scan signal controller comprises means for controlling the scan signals supplied to said first scan elements and said second scan elements to achieve a desired position of the scan origin for given parameter values of the charged particle beam. claim 14 18. Apparatus as defined in claim 14 wherein said scan signal controller comprises means for controlling the scan signals supplied to said first scan elements and said second scan elements to achieve a fixed position of the scan origin for different parameter values of the charged particle beam. claim 14 19. Apparatus as defined in claim 14 wherein said scan signal controller comprises means for controlling the scan signals supplied to said first scan elements and said second scan elements to achieve a desired position of the scan origin as the energy of the charged particle beam changes. claim 14 20. Apparatus as defined in claim 14 wherein said scan signal controller comprises means for controlling the scan signals supplied to said first scan elements and said second scan elements to change the effective length of said first scan elements and said second scan elements. claim 14 21. Apparatus as defined in claim 14 wherein said scan signal controller comprises means for adjusting the ratio of scan signals supplied to said first scan elements and said second scan elements. claim 14 22. Apparatus as defined in claim 14 further comprising a position controller for positioning one or both of said first scan elements and said second scan elements based on said at least one parameter of the charged particle beam. claim 14 23. Apparatus as defined in claim 14 wherein said scan signal controller comprises means for applying the scan signals to the first and second scan elements for scanning a high energy charged particle beam and means for applying the scan signals to the first scan plates and for grounding the second scan plates for scanning a low energy charged particle beam, wherein an effective length over which electric fields are applied to the charged particle beam is reduced for scanning a low energy beam. claim 14 24. Apparatus as defined in claim 23 wherein the charged particle beam comprises a positive ion beam and wherein the scan signal generator applies only negative voltages or ground to the first and second scan elements for scanning a low energy beam. claim 23 25. A method for scanning a charged particle beam, comprising the steps of: directing a charged particle beam between spaced-apart scan elements; energizing said scan elements for scanning the charged particle beam in a scan pattern having a scan origin; and s controlling positions of said scan elements based on at least one parameter of the charged particle beam. 26. A method as defined in claim 25 wherein the step of directing a charged particle beam comprises directing a charged particle beam between spaced-apart electrostatic scan plates and wherein the step of energizing said scan elements comprises coupling scan voltages to said scan plates. claim 25 27. A method as defined in claim 25 wherein the step of directing a charged particle beam comprises directing a charged particle beam between polepieces of a scan magnet and wherein the step of energizing said scan elements comprises coupling scan currents to a magnet coil of the scan magnet. claim 25 28. A method as defined in claim 25 wherein the step of positioning said scan elements comprises positioning said scan elements to achieve a fixed position of the scan origin for different parameter values of the charged particle beam. claim 25 29. A method as defined in claim 25 wherein the step of positioning said scan elements comprises positioning said scan elements to achieve a fixed position of the s can origin for different energies of the charged particle beam. claim 25 30. A method as defined in claim 26 wherein the step of positioning said scan elements comprises moving said scan plates upstream with respect to the charged particle beam and increasing the spacing between the scan plates. claim 26 31. A method as defined in claim 26 wherein the step of positioning said scan elements comprises varying the gap between said scan plates and rotating said scan plates. claim 26 32. A method as defined in claim 25 wherein the step of positioning said scan elements comprises positioning said scan elements to achieve a desired position of the scan origin for given parameter values of the charged particle beam. claim 25 33. A method as defined in claim 26 wherein the step of positioning said scan elements comprises moving said scan plates to two or more discrete positions based on the energy of the charged particle beam. claim 26 34. A method as defined in claim 26 wherein the step of positioning said scan elements comprises moving said scan plates along a continuous range of positions based on the energy of the charged particle beam. claim 26 35. A method for scanning a charged particle beam, comprising the steps of: directing a charged particle beam between spaced-apart first scan elements and spaced-apart second scan elements; applying scan signals to said first scan elements and said second scan elements for scanning the charged particle beam in a scan pattern having a scan origin; and controlling the scan signals applied to said first scan elements and said second scan elements based on at least one parameter of the charged particle beam. 36. A method as defined in claim 35 wherein the step of controlling the scan signals comprises controlling the scan signals applied to said first scan elements and said second scan elements to achieve a desired position of the scan origin for given parameter values of the charged particle beam. claim 35 37. A method as defined in claim 35 further comprising the step of controlling positions of one or both of said first scan elements and said second scan elements based on said at least one parameter of the charged particle beam. claim 35 38. Apparatus for scanning an ion beam, comprising: two or more pairs of scan plates for scanning the ion beam; and a scan generator for applying scan voltages to said two or more pairs of scan plates for scanning a high energy ion beam and for applying scan voltages to a subset of said two or more sets of scan plates for scanning a low energy ion beam, wherein unused scan plates are electrically grounded and wherein an effective length over which electric fields are applied to the ion beam is reduced for scanning a low energy beam. 39. Apparatus as defined in claim 38 wherein the ion beam is a positive ion beam and wherein said scan generator applies only negative voltages or ground to said two or more pairs of scan plates for scanning a low energy ion beam. claim 38 40. Apparatus as defined in claim 38 further comprising a position controller for positioning said two or more pairs of scan plates based on the energy of the ion beam. claim 38
	summary
047675934	abstract	A multiple shell pressure vessel utilizes a number of spaced apart, concentrically disposed pressure vessel shells within and spaced apart from an outer pressure vessel shell, the spaces between each shell being filled with a low melting point, high boiling point material selected from the group, lead, tin, antimony, bismuth, or sodium and potassium, and mixtures thereof and subjected to a pressure whereby the innermost pressure vessel shell wall is maintained in compression while the main vessel is pressurized. Chemical compositions or compounds which contain boron or cadmium may also be added to the filler material. The vessel may include devices for keeping the pressure in the space between the innermost pressure vessel shell and the next innermost shell at a fixed predetermined multiple of the pressure within the innermost pressure vessel shell. A method of constructing the multiple shell pressure vessel includes a sequence of steps for filling the spaces between shells with the low melting point, high boiling point materials in such a way that the inner shell is compressed and the outer shells are in tension.
056365128	summary	FIELD OF THE INVENTION This invention relates to nuclear rocket engines generally and more specifically to a nuclear rocket feed system that incorporates an auxiliary power cycle. BACKGROUND OF THE INVENTION Conventional nuclear rocket engines employ a nuclear fission reactor to heat the rocket propellant, typically hydrogen, to extremely large temperatures. The hot hydrogen is then expelled from a nozzle into space at supersonic speed to create thrust for the rocket. To conserve nuclear fuel and propellant, space mission operations will typically only require short duration engine firings. The reactor is turned on for a brief period to generate thrust to propel the rocket to a cruise velocity in space and then the reactor is shut down. Shutting down nuclear rocket engines during the space mission has presented many design challenges. One challenge results from the fact that nuclear reactors cannot be immediately turned off. Delay neutrons and daughter products of the fission reaction generate power long after the reactor ceases to operate. This energy or heat must be removed from the rocket to prevent overheating and destruction of the engine. In addition, the engine feed system (pumps and turbines) must be shut down as the reactor power decays to throttle propellant flow and to prevent the pumps from surging. Shutting down the feed system, however, makes it makes it extremely difficult to remove the reactor heat from the engine. Another challenge created by shutting down rocket engines in flight is cooling the nuclear reactor. Byproducts of the nuclear reaction (waste heat) continuously heat the components of the reactor during engine firing and long after the engine has been shut down. To solve this problem, liquid or gaseous hydrogen (apart from the actual propellant) is typically used to cool the reactor. The hydrogen is directed through the reactor, which transfers some of its heat to the hydrogen, and is then expelled from the rocket into space. This process continues until the reactor temperature has been brought down to a safe level. One problem with this method is that cooling the reactor can take a long time (from a few hours to a few days). Thus, an enormous amount of hydrogen must be stored in the rocket to cool the reactor. This large volume of hydrogen increases the weight of the rocket which decreases mission performance (payload/initial mass) and increases mission cost. To decrease the amount of hydrogen needed to cool the reactor, existing systems have attempted to alternate between undercooling and overcooling the reactor. In this scheme, the reactor is allowed to heat up until it reaches a very high temperature and is then quickly cooled down with extremely cold propellant. Hydrogen is conserved because it is not continuously pumped through the reactor and into space. Alternating between undercooling and overcooling the reactor, however, can create thermal shocks that damage reactor components and create flow instabilities, thereby decreasing the life of the nuclear engine. SUMMARY OF THE INVENTION The present invention solves these problems of shutting down the engine during flight and cooling the reactor throughout the space mission. To accomplish this, a nuclear rocket engine includes a primary feed system for pumping rocket propellant from a propellant source to a nuclear reactor and an auxiliary feed system coupled to the primary feed system. The auxiliary feed system can be configured into a high thrust mode for withdrawing heat from the engine when the reactor is operating at full power, a low thrust mode for throttling propellant flow and radiating heat into space during reactor shutdown and a zero thrust mode for cooling the nuclear reactor and generating electricity for the rocket's auxiliary power requirements during the remainder of the mission. In the high thrust mode, the auxiliary feed system includes a bypass line with an inlet coupled to a recycling port in the primary feed system, an outlet coupled to the propellant source and means for withdrawing heat from propellant flowing along the bypass line. A recuperator is coupled to the primary feed system for transferring heat from the hot propellant in the reactor to the cool propellant from the propellant source. A portion of this now heated propellant is bled into the auxiliary feed system after passing through the recuperator to withdraw heat from the engine. This heat can be converted into electricity to power other operations on the rocket or discharged into space to release heat from the engine or control the attitude of the rocket. Preferably, the auxiliary feed system is a Brayton power cycle having a turbine coupled to a compressor for pumping propellant through the feed system. The heat withdrawing means comprises a space radiator for withdrawing heat from the warm propellant that has passed through the recuperator and discharging this heat into space. The heat withdrawing means may also include a motorgenerator coupled to the turbine for translating the mechanical energy of the turbine into electricity. The cooled gaseous hydrogen is then recycled back into the propellant source to maintain a suitable pressure within the propellant source. In the low thrust mode, the auxiliary feed system has an inlet coupled to a recycling port in the nuclear reactor between the fuel assemblies and the nozzle and an outlet coupled to the reactor between the reactor inlet and the fuel assemblies. In this mode, a portion of the hot propellant exiting the fuel assemblies is bled into the auxiliary feed system to discharge heat (generated by the neutron delay reactions in the fuel assemblies) into space. Some of this heat is used to drive the turbine so that the auxiliary feed system can pump the propellant through the primary feed system. In this manner, the main pumps and turbines in the primary feed system can be shut down to facilitate throttling of the propellant flow and to prevent the pumps from surging during reactor shutdown. Similar to the high thrust mode, electricity is generated from the waste heat with the motorgenerator and the cooled propellant maintains a suitable pressure in the propellant source. In the zero thrust mode, the auxiliary feed system has an inlet coupled to the reactor between the fuel assemblies and the nozzle and an outlet coupled to the primary feed system. In this mode, the nozzle is bypassed so that all of the hot propellant from the fuel assemblies flows into the auxiliary power system to discharge heat through the space radiator. The cooled hydrogen is then recycled back through the reactor to continue the cooling process. Thus, the reactor can be completely cooled without losing any hydrogen. This results in a substantial decrease in the amount of hydrogen needed for a space mission and, therefore, a substantial decrease in the weight of the rocket. Another advantage of the present invention is that reactor coolant gas is preheated before cooling the fuel assemblies in the low and zero thrust modes. Preheating the reactor coolant gas increases flow stability and proper cooling of the fuel assemblies even at low flow velocities. This increases the life of the system because thermal shocks and flow instabilities are avoided. The above is a brief description of some deficiencies in the prior art and advantages of the present invention. Other features, advantages and embodiments of the invention will be apparent to those skilled in the art from the following description, accompanying drawings and appended claims.
	summary
052788873	claims	1. In an X-ray fluoroscopic apparatus for passing X-rays from an X-ray source to an X-ray detector, through a subject body; a radiation reduction device comprising: a filter member, being semi-transparent to X-Rays, and having at least one aperture formed therethrough, such that X-rays passing through said at least one aperture remain attenuated and strike said subject body in a common region, wherein the X-rays passing through said filter member are attenuated and strike said subject body in a pattern that surrounds, and is adjacent to, said common region; and means for increasing the image intensity of the attenuated X-rays to substantially correspond to the image intensity produced by said unattenuated X-rays. selectively filtering said X-ray radiation such that attenuated and unattenuated X-ray radiation strike said subject body, said unattenuated X-ray radiation being confined to a predetermined common area surrounded by said attenuated X-ray radiation; and increasing the intensity of an image produced by said attenuated X-radiation to substantially correspond to the intensity of an image produced by said unattenuated X-ray radiation. a controllable filter member means being responsive to an operation control signal, and including a filter member being semi-transparent to X-rays and having at least one aperture formed therethrough, such that X-rays passing through said at least one aperture remain unattenuated and strike said subject body in a common region; wherein an image is formed in which the X-rays passing through said filter member are attenuated and strike said subject body in an attenuated pattern that surrounds, and is adjacent to, said common region, with an annular transition region between said attenuated pattern and said common region; a method for compensating for differences in image intensity in said transition region, comprising the following steps: a. calculating an average intensity of said given arcuate segment for said outside boundary of said transition region; b. calculating an average intensity of said given arcuate segment 42 for said inside boundary of said transition region; c. calculating an average intensity within each arcuate segment; and d. applying linear interpolation to said inside and outside boundary average intensity values to approximate what the average intensity ought to be for the full range of radii of said given arcuate segment; e. determining the difference between said overall average intensity of said given arcuate segment and the interpolated intensity, at a given radius to create a correct compensation factor for said given arcuate segment; f. adding said difference to said interpolated intensity and repeat for the entire range of radii within the arcuate segment; and g. repeating the foregoing for each of said arcuate segments into which said transition region 32 has been divided. selectively filtering said X-ray radiation such that attenuated and unattenuated X-ray radiation pass through said patient, said unattenuated X-ray radiation being confined to a common region; calculating the size and position of said common region striking said patient; altering the position of said common region to follow the advancement of said medical instrument, such that a point of interest on said medical instrument is viewed within said common region; and increasing the intensity of said viewed image produced by said attenuated X-ray radiation to substantially correspond to the intensity of said viewed image produced by said unattenuated X-ray radiation. a controllable filter member being responsive to an operation control signal, and including a filter member being semi-transparent X-rays and having at least one aperture formed therethrough, such that X-rays passing through said at least one aperture remain unattenuated and strike said subject body in a common region, wherein the X-rays passing through said filter member are attenuated and strike said subject body in a pattern that surrounds, and is adjacent to, said common region; control means for providing said operation signal, being coupled to said X-ray source means, to said image processing means, and to said controllable filter member means to control operation of said controllable filter member; and means for increasing image intensity of the attenuated X-rays to substantially correspond to image intensity produced by said unattenuated X-rays. a controllable filter member means being responsive to an operation control signal, and including a filter member being semi-transparent to X-rays and having at least one aperture formed therethrough, such that X-rays passing through said at least one aperture remain unattenuated and strike said subject body in a common region; wherein the X-rays passing through said filter member are attenuated and strike said subject body in an attenuated pattern that surrounds, and is adjacent to, said common region; control means coupled to said X-ray source means, to said image processing means, and to said controllable filter member means for providing said operation control signal to control operation of said controllable filter members, a method for correcting gray-scale values in said attenuated pattern region, comprising the following steps: 2. The device of claim 1, wherein said filter member includes a transition area surrounding said at least one aperture, said transition area having an increased transparency to X-rays as said transition area approaches said at least one aperture. 3. The device of claim 2, wherein said filter member is a substantially planar structure having a single aperture formed therethrough, said planar structure decreasing in thickness in said transition area such that the thickness of said planar structure is at a minimum at the edge of said aperture. 4. In an X-ray fluoroscopic procedure wherein an image is produced by passing X-ray radiation through a subject body a method of reducing the dosage of X-ray radiation striking said subject body, comprising the steps of: 5. The method of claim 4, wherein said step of increasing the intensity of said image includes dividing the area of said image, formed by X-ray radiation passing through said transition area, into a plurality of segments and increasing the intensity of each said segment to substantially correspond to the intensity of said common region. 6. The method of claim 4, further including the step of automatically positioning said common area over a point of interest in said subject body. 7. The method of claim 4, wherein said step of selectively filtering is accomplished by a filter member that is semi-transparent to X-ray radiation, said filter member having at least one aperture formed therethrough so that X-ray radiation can pass through said filter member, and into said common region, unattenuated. 8. The method of claim 7, wherein said filter member includes a transition area surrounding said at least one aperture, said transition area having an increased transparency to said X-ray radiation as said transition area approaches said at least one aperture. 9. The method of claim 8, wherein said filter member is a metal plate having a single aperture formed therethrough, said metal plate decreasing in thickness in said transition area such that the thickness of said metal plate is at a minimum at the edge of said aperture. 10. In an X-ray fluoroscopic apparatus for passing X-rays from an X-ray source means through a subject body to an X-Ray detector means, including image processing means including image intensifier means, coupled thereto for providing an image and including a radiation reduction device comprising: 3. calculating an intensity profile of said given arcuate segment; 11. The method recited in claim 10, wherein step 3 comprises: 12. In an X-ray fluoroscopic procedure wherein a viewed image is produced, for monitoring the advancement of a medical instrument within a patient, by passing X-ray radiation through a patient; a method for reducing the dosage of X-ray radiation being exposed to said patient comprising the steps of: 13. The method of claim 12, wherein said step of calculating the size and position of said common region includes obtaining a gradient image of said viewed image, calculating projections from said gradient image along a plurality of orientations, finding peaks in the data of said projections, and comparing said peaks in said projections determining the size and location of said common region. 14. The method of claim 13, wherein said step of calculating the size and position of said common region further includes smoothing said peaks to remove false peaks. 15. The method of claim 14, where four projections are calculated from said gradient image, said projections being taken at 0.degree., 45.degree., 90.degree. and 130.degree. relative to the horizontal. 16. The method of claim 15, wherein said step of selectively filtering is accomplished by a filter member that is semi-transparent to X-ray radiation, said filter member having at least one aperture formed therethrough so that X-ray radiation can pass through said filter member, and into said common region, unattenuated. 17. The method of claim 16, where said step of automatically altering the position of said common region includes locating said point of interest of said medical instrument in said viewed image in real time, relaying the position of said point of interest to a systems controller, using said systems controller to adjust the position of said filter member so that said common region corresponds in position to said point of interest. 18. The method of claim 17, wherein said filter member includes a transition area surrounding said at least one aperture, said transition area having an increased transparency to said X-ray radiation as said transition area approaches said at least one aperture. 19. The method of claim 18, wherein said filter member is a metal plate having a single aperture formed therethrough, said metal plate decreasing in thickness in said transition area such that the thickness of said metal plate is at a minimum at the edge of said aperture. 20. An X-Ray fluoroscopic apparatus for passing X-rays from an X-ray source means through a subject body to an X-ray detector means, including image processing means coupled thereto; a radiation reduction device comprising; 21. The apparatus recited in claim 20, wherein said control means, in response to an operative condition of said X-ray source means, acquires an image from said image processing means for calibration. 22. The apparatus recited in claim 20, wherein said X-ray source means includes means for providing a signal to said control means indicative of movement of said subject body. 23. The apparatus recited in claim 22, wherein said control means causes said X-ray source means to supply a dose of X-rays through said subject body in response to said signal indicative of movement. 24. The apparatus recited in claim 20, wherein said image processing means includes image intensifier means, said image processing means providing a signal to said control means indicative of a change of distance between said image intensifier means and said subject body, whereby a change of geometric magnification results. 25. The apparatus recited in claim 20, wherein said control means provides said control signal to said controllable filter member means for selectably placing said filter member in an operative mode. 26. In an X-ray fluoroscopic apparatus for passing X-rays from an X-ray source means through a subject body to an X-ray detector means, including image processing means including image intensifier means, coupled thereto for providing an image and including a radiation reduction device comprising:
047160136	claims	1. A nuclear reactor including a vessel, a nuclear core within said vessel, control-rod assemblies within said vessel, said control-rod assemblies including vertical control-rod guide means in a plenum vertically above said core, said control-rod assemblies also including control-rod clusters, each said cluster including a plurality of control rods suspended from a spider, said cluster being movable to move its said control rods between said guide means and said core, and a drive rod connected to each said cluster for so moving said each cluster, said vessel having at least an inlet nozzle for supplying a coolant through said core, the inflowing coolant after passing through said core flowing predominantly through said guide means generally vertically, a calandria in the path of said outflowing coolant, from said guide means, said calandria including a plurality of hollow members, an upper support for said members perforate to the opening within said members and a lower support, said calandria being mounted with its lower support above and on said guide means and with said drive rods only and not said control rods passing through said hollow members, said lower support being perforate to the coolant flowing out of said guide means, said vessel having at least an outlet nozzle with a generally horizontal coolant outflow channel, said outflow channel being substantially at the level of said calandria so that the coolant flowing into said calandria flows generally transversely over the outer surfaces of said hollow members and out through said outflow channel. 2. The reactor of claim 1 wherein the hollow members are so constructed that the stresses in said hollow members resulting from the flow of coolant transversely through said members are substantially below the level at which failure of the hollow members results from the transverse flow of the coolant through them at a high rate. 3. A nuclear reactor including a vessel, a nuclear core within said vessel, control-rod assemblies within said vessel, said control-rod assemblies including vertical control-rod guide means in a plenum vertically above said core, said control-rod assemblies also including control-rod clusters, each said cluster including a plurality of control rods suspended from spiders, said cluster being movable to move its said control rods between said guide means and said core, and a drive rod connected to each said cluster for so moving each said cluster, said vessel having at least an inlet nozzle for supplying a coolant through said core, the inflowing coolant after passing through said core flowing predominantly through said guide means generally vertically, a calandria in the path of said outflowing coolant, said calandria including a plurality of hollow members, an upper support for said members perforate to the opening within said members and a lower support, said calandria being mounted with its lower support above and on said guide means and with only said drive rods and not said control rods passing through said hollow members, said lower support being perforate to the coolant flowing out of said guide means, said vessel having at least an outlet nozzle with a generally horizontal coolant outflow channel, said coolant flowing into the calandria flowing generally transversely over the outer surfaces of said hollow members and out through said outflow channel, said outflow channel being so spaced with respect to said calandria that there is no substantial pressure drop between the coolant flowing out of the calandria and the coolant flowing into said outflow channel. 4. A nuclear reactor including a vessel, a nuclear core within said vessel, control-rod assemblies within said vessel, said control-rod assemblies including vertical control-rod guide means in a plenum vertically above said core, said control-rod assemblies also including control-rod clusters, each said cluster including a plurality of control rods suspended from a spider, each said cluster being movable to move its said control rods, between said guide means and said core, and a drive rod connected to each said cluster for so moving said each said cluster, said vessel having at least an inlet nozzle and an outlet nozzle with a generally horizontal coolant outflow channel for circulating a coolant through said core, the inflowing coolant after passing through said core flowing predominantly through said guide means generally vertically, said outflow channel being substantially at a level just above the upper end of said guide means, a calandria in the path of said outflowing coolant, said calandria including a plurality of hollow members, an upper support for said members perforate to the opening within said members and a lower support, said calandria being mounted with its lower support above and on said guide means and with only said drive rods and not said control rods passing through said hollow members, said lower support being perforate to the coolant flowing out of said guide means, the coolant flowing out of said guides flowing directly out of said outflow channel passing generally transversely over the outer surfaces of said hollow members. 5. The nuclear reactor of claim 3 wherein the vessel includes a generally cylindrical body and a head, the calandria includes a supporting flanged cylinder extending from the top thereof, and the flange of the cylinder is secured between the head and the body. 6. The nuclear reactor of claim 3 wherein the vessel includes a generally cylindrical body and a head, said body having an internally extending lip, and wherein the guide means, the core, and a substantial portion of the calandria are encircled by a first barrel having a first flange and the guide means and the calandria are encircled by a second barrel having a second flange and the calandria has a cylinder having a third flange extending from the top thereof, the first and second barrels and the cylinder being mounted generally coaxially with the first, second and third flanges compressed between the lip and the head. 7. The nuclear reactor of claim 6 wherein the first barrel and the body of the vessel define an annulus communicating with the inflow channel of the inlet nozzle through which the inflowing coolant is conducted, there being means preventing the flow of coolant into the outflowing channel between the annulus and the outlet nozzle. 8. The nuclear reactor of claim 7 wherein the flow channel of the inlet nozzle is substantially at the same level as the flow channel of the outlet nozzle. 9. The nuclear reactor of claim 8 wherein the length of the portions of said hollow members between the upper and lower support is at least equal to the cross dimension of the outlet nozzle along the hollow members. 10. The nuclear reactor of claim 3 wherein the volume of the calandria between the upper and lower supports in such that the pressure distribution in the coolant within the calandria is substantially uniform whereby the flow of coolant in the calandria does not produce materially non-uniform pressure distribution in the core. 11. The nuclear reactor of claim 9 wherein the inlet nozzle is also at the level of the calandria. 12. The nuclear reactor of claim 11 wherein the guides of the upper internals are enclosed by a shell having former plates spaced along the length thereof, said former plates extending toward the periphery of said guides, there being a small gap between each of said former plates and said periphery along which coolant flows upwardly, whereby the pressure of said coolant within and surrounding said guides is equalized. 13. The nuclear reactor of claim 12 wherein the walls of the guides are substantially imperforate to coolant which would flow laterally out of said guides so as to suppress substantial flow of coolant laterally out of said guides but each said guides is perforate at its lower and upper ends to permit the inflow of coolant vertically from said core into said each guide and the outflow generally vertically of the coolant from said guide means. 14. The nuclear reactor of claim 3 wherein the walls of the guide means are substantially imperforate to coolant latterally out of said guide means so as to suppress substantial flow of coolant laterally out of said guide means but said guide means are perforate at its lower and upper ends to permit the inflow of coolant vertically from said core into said guide means and the outflow vertically of the coolant from said guide means. 15. The nuclear reactor of claim 4 wherein the walls of the guide means are substantially imperforate to coolant laterally out of said guide means so as to suppress substantial flow of coolant laterally out of said guide means but said guide means are perforate at its lower and upper ends to permit the inflow of coolant vertically from said core into said guide means and the outflow vertically of the coolant from said guide means. 16. A nuclear reactor including a vessel, a nuclear core within said vessel, control-rod clusters each cluster including a plurality of control rods for controlling said reactor suspended from a spider, upper internals mounted on said core vertically above said core, said upper internals including guides, a calandria having a plurality of hollow members mounted between upper and lower supports perforate to said hollow members, said calandria being mounted on said upper internals vertically above said upper internals, with said lower support on said upper internals, a drive rod, connected to each said cluster, each said drive rod only and not said control rods passing moveably through a said hollow member of said calandria and suspending the cluster connected to it movable by said drive rod between a position in which, said control rod clusters are within said guides and said spider is near said lower support and a position in which the spider of said each cluster is near said core and said control rods are within said core, said vessel having at least an inlet nozzle and an outlet nozzle, said outlet-nozzle having a generally-horizontal coolant outflow channel, said outflow channel being substantially at the level of said hollow members, and means for conducting coolant from said inlet nozzle in succession through said core, said upper internals, said hollow members and said outflow channel, said coolant being conducted generally vertically through said core and upper internals, generally transversely through said hollow members and generally horizontally through said outflow channel. 17. A nuclear reactor including a vessel, a nuclear core within said vessel, control-rod clusters each cluster including a plurality of control rods for controlling said reactor suspended from a spider, upper internals mounted on said core vertically above said core, said upper internals including guides, a plurality of hollow members mounted on said upper internals vertically above said upper internals, a drive rod, connected to each said cluster, only each said drive rod and not said control rods passing moveably through a said hollow member and suspending the cluster connected to it movable by said drive rod between a position in which, said control rod clusters are within said guides and said spider is near said lower support and a position in which the spider of said each cluster is near said core and said control rods are within said core, said vessel having at least an inlet nozzle and an outlet nozzle, said outlet-nozzle having a generally-horizontal coolant outflow channel, said outflow channel being substantially at the level of said hollow members, and means for conducting coolant from said inlet nozzle in succession through said core, said upper internals, said hollow members and said outflow channel, said coolant being conducted generally vertically through said core and upper internals, generally transversely through said hollow members and generally horizontally through said outflow channel. 18. A nuclear reactor including a vessel, a nuclear core within said vessel, a plurality of control rods for controlling said reactor, upper internals mounted on said core vertically above said core, said upper internals including guides for said control-rod rods, a plurality of generally vertical hollow members mounted on said upper internals vertically above said upper internals, a plurality of drive rods connected to said plurality of control rods, each drive rod connected to certain of the control rods of said plurality of control rods, only said drive rods and not said control rods passing moveably through said hollow member and suspending said certain of said control rods connected to said drive rods movable by said drive rods between a position in which said control rods are within said guides and a position in which said control rods are wholly within said core, said vessel having at least an inlet nozzle and an outlet nozzle, said outlet nozzle having a generally horizontal coolant outflow channel, said outflow channel being substantially at the level of said hollow members, and means for conducting coolant from said inlet nozzle in succession through said core, said upper internals, said hollow members and said outflow channel, said coolant being conducted generally vertically through said core and said upper internals, generally transversely through said hollow members and generally horizontally through said outflow channel. 19. A nuclear reactor including a vessel, a nuclear core within said vessel, control rods for controlling said core, control-rod guide means mounted within said vessel vertically above said core, drive rods, connected to said control rods, for driving said control rods between said control-rod guide means and said core, a plurality of generally vertical tubes mounting vertically above said guide means within said vessel, only said drive rods and not said control rods passing moveably through said tubes, means for conducting a coolant through said vessel, said conducting means including at least an inlet nozzle and an outlet nozzle, said coolant being conducted in the normal operation of said reactor into said vessel through said inlet nozzle and then generally vertically upwardly in succession through said core and said control-rod guide means and then generally horizontally past said tubes and out through said outlet nozzle, said outlet nozzle being at the level of said tubes and subtending said tubes along the predominant portion of their lengths.
	description	This application claims the benefit of Korean Patent Application No. 10-2016-0008885, filed on Jan. 25, 2016, in the Korean Intellectual Property Office, the disclosure of which is incorporated herein in its entirety by reference. 1. Field One or more embodiments relate to a sealing mechanism for a small-medium reactor vessel (RV) cable penetration tube, and more particularly, to a sealing mechanism for a small-medium RV cable penetration tube, the sealing mechanism being configured to improve the functional and structural integrity of a cable inserted into a small-medium RV through a penetration tube by using a precise thimble. 2. Description of the Related Art In general, control devices are installed in small and medium reactor vessels (RVs) to control nuclear reactors. Control cables for operating such control devices are inserted from the outside to the inside of RVs by using tubes penetrating the RVs, and sealing mechanisms are used to prevent leakage of cooling materials from nuclear reactors. Control cables inserted into small and medium RVs are required to operate normally under high-temperature, high-pressure, high-radioactivity, subaqueous conditions and secure sealing mechanisms have to be used to prevent leakage of cooling materials from nuclear reactors along cables arranged in penetration tubes. Particularly, control cables through which power and various signals are provided to control rod driving devices installed in RVs are required to operate without replacement for a given period of time for safe operations of nuclear reactors. Although such control cables are required to have functional and structural integrity under high-temperature, high-pressure, high-radioactivity, subaqueous conditions, the current technology for improving the integrity of cables is insufficient. FIG. 1 is a view illustrating a cable 30 inserted in an RV 10 through a penetration tube 20 in the related art. Referring to FIG. 1, since a device or structure for supporting the cable 30 is not provided in the related art, the cable 30 may wobble due to fluid flows or various internal vibrations occurring in the RV 10. Therefore, the structural integrity of the cable 30 may deteriorate. Fluid flows and various internal vibrations occurring in the RV 10 may damage the cable 30 in addition to lowering the structural integrity of the cable 30. Thus, the cable 30 may not function normally and the functional integrity of the cable 30 may deteriorate. One or more embodiments include a sealing mechanism for a small-medium reactor vessel (RV) cable penetration tube, the sealing mechanism being configured to improve the functional and structural integrity of a cable inserted into a small-medium RV through a penetration tube by using a precise thimble. Additional aspects will be set forth in part in the description which follows and, in part, will be apparent from the description, or may be learned by practice of the presented embodiments. According to one or more embodiments, there is provided a sealing mechanism for a small-medium RV cable penetration tube, the sealing mechanism including: a penetration tube configured to penetrate an RV from an outside to an inside thereof and including a penetration hole for communication with the inside of the RV; a cable configured to be inserted in the RV through the penetration hole of the penetration tube; and a thimble placed between the cable and the penetration tube, wherein a dimple groove portion is provided on the thimble in a direction from an outer surface of the penetration hole toward the cable. The thimble may have a tube shape and may be placed between the cable and the penetration tube, the dimple groove portion may extend along an outer surface of the thimble, and an end of the dimple groove portion and another end of the dimple groove portion may make an angle of about 120° to about 180° with a center of the thimble. The dimple groove portion may have a round groove shape. The sealing mechanism may further include a guide tube and a Swagelok fitting, wherein the guide tube may protrude from an end of the penetration tube and may include a guide hole communicating with the penetration hole to receive the cable, and the Swagelok fitting may be installed on the end of the penetration tube or an end of the guide tube. The diameter of the penetration tube may be smaller at an end portion and another end portion of the penetration tube than at a middle portion of the penetration tube, and the thimble may be supported by the end portion and the other end portion of the penetration tube. The thimble may include a thimble protrusion between the end portion and the middle portion of the penetration tube or between the other end portion and the middle portion of the penetration tube. The thimble may be shorter than the penetration hole such that an interval may be defined between the thimble and an end of the penetration tube or between the thimble and another end of the penetration tube. The penetration tube may include at least one penetration hole, and the penetration tube may be provided in a lateral side of the RV. Reference will now be made in detail to embodiments, examples of which are illustrated in the accompanying drawings, wherein like reference numerals refer to like elements throughout. In this regard, the present embodiments may have different forms and should not be construed as being limited to the descriptions set forth herein. Accordingly, the example embodiments are merely described below, by referring to the figures, to explain aspects of the present description. As used herein, the term “and/or” includes any and all combinations of one or more of the associated list. The example embodiments relate to a sealing mechanism for a small-medium reactor vessel (RV) cable penetration tube configured to improve the functional and structural integrity of a cable inserted into a small-medium RV through a penetration tube due to use of a precise thimble. Hereinafter, the example embodiments will be described with reference to the accompanying drawings. Referring to FIG. 2, according to an example embodiment, a sealing mechanism for a small-medium RV cable penetration tube includes a penetration tube 120, a cable 130 (two are shown), a thimble 140 (two are shown), and a dimple groove portion 141 (two or more are shown). The penetration tube 120 penetrates an RV 110 from the outside to the inside thereof and includes a penetration hole 121 for communication with the inside of the RV 110. The penetration hole 121 extends from an end of the penetration tube 120 to another end of the penetration tube 120 and communicates with the inside of the RV 110. The cable 130 may be introduced into the RV 110 by inserting the cable 130 into the penetration hole 121 of the penetration tube 120. The cable 130 may include various kinds of cables. Examples of the cable 130 may include a control cable connected to a controller placed in a small-medium RV and a cable connected to an in-reactor instrument configured to measure the neutron flux distribution and temperature of a reactor core of an RV. Besides the above-mentioned cables, examples of the cable 110 may include any kind cables insertable into the RV 110. The thimble 140 is placed between the cable 130 and the penetration tube 120. The thimble 140 may have a thin tube shape and surround the cable 130. The thimble 140 may include a zircalloy. The thimble 140 may surround the cable 130 and may function as a structural support for the cable 130. The shape of the thimble 140 is not limited to a tube shape. The thimble 140 may have any shape as long as the thimble 140 functions as a structural support for the cable 130. The dimple groove portion 141 includes a groove formed in a direction from an outer surface of the penetration hole 121 toward the cable 130. That is, the dimple groove portion 141 may have a groove shape recessed into the thimble 140 toward the cable 130. The dimple groove portion 141 may be in just contact with an outer surface of the cable 130. If the dimple groove portion 141 is in just contact with the cable 130, wobbling of the cable 130 may be prevented, and the structural integrity of the cable 130 may be improved (however, the dimple groove portion 141 may not be completely in just contact with the cable 130 but may be separate from the cable 130 by a slight distance). The dimple groove portion 141 may extend in an outer surface of the thimble 140 having a tube shape, and an end and another end of the dimple groove portion 141 may make an angle of about 120° to about 180° with the center of the thimble 140. That is, the thimble 140 may have a circular cross-sectional shape, and the center of the thimble 140 may refer to the center of the circular cross-sectional shape. The dimple groove portion 141 may extend within an angle range of about 120° to about 180° along the outer surface of the thimble 140 having a 360° circular cross-sectional shape. The dimple groove portion 141 may extend more than about 180°, or less than about 120°. However, if the dimple groove portion 141 extends more than about 180°, a contact area between the dimple groove portion 141 and the cable 130 may increase, and thus the magnitude of force acting on the cable 130 may increase. On the contrary, if the dimple groove portion 141 extends less than about 120°, the cable 130 may be poorly supported. Therefore, the end and the other end of the dimple groove portion 141 may make an angle of about 120° to about 180° with the center of the thimble 140. Referring to FIG. 3, a plurality of dimple groove portions 141 are formed in a length direction of the thimble 140. The dimple groove portions 141 may be formed at regular or irregular intervals in the length direction of the thimble 140. In addition, the dimple groove portions 141 may be formed in a side-to-side alternating manner in the length direction of the thimble 140. If the dimple groove portions 141 are densely formed in a particular region of the thimble 140, a large force may be applied to a particular position of the cable 130, and thus negative effects may arise. The dimple groove portions 141 may have a round groove shape. The dimple groove portions 141 may be in contact with the cable 130. Therefore, if the dimple groove portions 141 have a sharp shape, the cable 130 may be damaged by the dimple groove portions 141. Thus, the dimple groove portions 141 may have a round shape. However, the shape of the dimple groove portions 141 is not limited a round shape. That is, the dimple groove portions 141 may have any shape as long as the dimple groove portions 141 support the cable 130 without damaging the cable 130. In other words, the dimple groove portions 141 are not limited to the above-mentioned shape but may have various shapes. For example, the dimple groove portions 141 may not have a groove shape extending along the outer surface of the thimble 140 but may be formed in the shape of grooves not extending along the outer surface of the thimble 140. In addition, the dimple groove portions 141 may have any other shape capable of supporting the cable 130. According to an embodiment, the penetration tube 120 of the sealing mechanism for a small-medium RV cable penetration tube may include a penetration tube end portion 122, a penetration tube middle portion 123, and another penetration tube end portion 124. The penetration tube end portion 122 may be located outside the RV 110, and the other penetration tube end portion 124 may be located inside the RV 110. Referring to FIG. 2, the diameter of the penetration hole 121 may be smaller at the penetration tube end portion 122 and the other penetration tube end portion 124 than at the penetration tube middle portion 123. That is, the penetration tube end portion 122 and the other penetration tube end portion 124 may be narrower than the penetration tube middle portion 123. At least one penetration hole 121 may be formed in the penetration tube 120. If a plurality of penetration holes 121 are formed in the penetration tube 120, a plurality of cables 130 may be inserted into the RV 110 through the penetration tube 120. When a plurality of penetration holes 121 are formed in the penetration tube 120, the plurality of penetration holes 121 may communicate with the penetration tube middle portion 123. That is, the plurality of penetration holes 121 may not communicate with each other in the penetration tube end portion 122 and the other penetration tube end portion 124 but may communicate with each other in the penetration tube middle portion 123. The thimble 140 is placed around the cable 130 in the penetration hole 121 and may be inserted into and supported by the penetration tube end portion 122 and the other penetration tube end portion 124. Referring to FIG. 2, the diameter of the penetration hole 121 at the penetration tube end portion 122 and the other penetration tube end portion 124 may be equal to or slightly greater than the diameter of the thimble 140 having a tube shape. Owing to this structure, an end of the thimble 140 may be inserted into the penetration tube end portion 122, and another end of the thimble 140 may be inserted into the other penetration tube end portion 124. Since the diameter of the penetration hole 121 at the penetration tube end portion 122 and the other penetration tube end portion 124 is equal to or slightly greater than the diameter of the thimble 140, the thimble 140 may be supported without wobbling by the penetration tube end portion 122 and the other penetration tube end portion 124. Thus, the thimble 140 may be supported without being shaken by fluid flows or other internal vibrations occurring in the RV 110. Since the thimble 140 is supported as described above, the cable 130 placed inside the thimble 140 may be supported without being shaken by fluid flows or other internal vibrations occurring in the RV 110, and thus the structural integrity of the cable 130 may be improved. In addition, the diameter of the thimble 140 at the penetration tube end portion 122 and the other penetration tube end portion 124 is equal to or slightly greater than the diameter of the cable 130, and thus fine gaps may be formed between the cable 130 and the thimble 140. Such fine gaps may generate a pressure difference with respect to a fluid contained in the RV 110. The thimble 140 inserted into the penetration tube end portion 122 and the other penetration tube end portion 124 may be shorter than the penetration hole 121, and thus there may be an interval between the thimble 140 and an end of the penetration tube 120 or another end of the penetration tube 120. The penetration hole 121 extends from the end of the penetration tube 120 to the other end of the penetration tube 120, and the thimble 140 is shorter than the penetration hole 121. Therefore, a thermal expansion space 125 may be formed between the thimble 140 and the end or the other end of the penetration tube 120. Referring to FIG. 3, the thermal expansion space 125 may be formed in the penetration tube end portion 122 in a region defined from the end of the thimble 140 to the end of the penetration tube 120. Owing to the thermal expansion space 125, the thimble 140 may not be damaged during thermal expansion. If the thermal expansion space 125 is not formed, the thimble 140 may be damaged because there is no space for accommodating thermal expansion of the thimble 140. That is, since the thimble 140 is shorter than the penetration hole 121, the thermal expansion space 125 is formed, and thus the thimble 140 may not be damaged by heat as well as the cable 130 may be supported safely. The position of the thermal expansion space 125 is not limited to the penetration tube end portion 122. For example, the thermal expansion space 125 may be formed in the other penetration tube end portion 124. In addition, any other space may be formed to prevent the thimble 140 from being damaged by thermal expansion. Referring to FIG. 2, a dimple protrusion 142 may be provided on the thimble 140 between the penetration tube end portion 122 and the penetration tube middle portion 123 or between the other penetration tube end portion 124 and the penetration tube middle portion 123. The dimple protrusion 142 may prevent separation of the thimble 140 from the penetration tube 120. That is, the diameter of the thimble 140 in a boundary region between the other penetration tube end portion 124 and the penetration tube middle portion 123 is greater than the diameter of the penetration hole 121 in the other penetration tube end portion 124. Since the thimble 140 have a diameter greater than the diameter of the penetration hole 121 formed in the other penetration tube end portion 124 as described above, the thimble 140 may be caught on the other penetration tube end portion 124, and thus the thimble 140 may not be separated from the penetration tube 120. As long as the dimple protrusion 142 prevents the thimble 140 from being separated from the penetration tube 120, the dimple protrusion 142 may be formed in a region between the penetration tube end portion 122 and the penetration tube middle portion 123 or in a region between the other penetration tube end portion 124 and the penetration tube middle portion 123. The dimple protrusion 142 may be continuously formed along the outer surface of the thimble 140 or may include one or more protrusions independently formed on the outer surface of the thimble 140. The dimple protrusion 142 is not limited to the above-described shape. The dimple protrusion 142 may have any shape as long as the dimple protrusion 142 prevents separation of the thimble 140 from the penetration tube 120. A guide tube 150 may protrude from the end of the penetration tube 120, and a guide hole 151 may be formed in the guide tube 150 in communication with the penetration hole 121 to receive the cable 130. The guide tube 150 is provided to easily insert the cable 130 into the penetration tube 120. That is, the cable 130 may be inserted into the penetration tube 120 while being guided by the guide tube 150. Referring to FIG. 4, a Swagelok fitting 160 may be coupled to the end of the penetration tube 120 so as to finally seal the penetration tube 120. The Swagelok fitting 160 is a sealing device configured to provide sealing by pushing front edges of ferrules 161 having a cone ring shape using wedge-shaped push rings. The Swagelok fitting 160 is a well-known device, and thus a detailed description thereof will not be presented here. The Swagelok fitting 160 may be directly coupled to the end of the penetration tube 120, or after the guide tube 150 is provided on the end of the penetration tube 120, the Swagelok fitting 160 may be coupled to the guide tube 150. The penetration tube 120 may be provided in a lateral side of the RV 110. Due to physical shape characteristics of the RV 110 related to relevant facilities such as a vapor generator connected to the RV 110, the penetration tube 120 for the cable 130 may be provided in a lateral side of the RV 110 rather than in an upper side of the RV 110 so as to improve the efficiency of subsequent processes. That is, many other structures may be installed on the upper side of the RV 110. Therefore, the penetration tube 120 may be installed in the lateral side of the RV 110 to avoid interference with other structures. The sealing mechanism for a small-medium RV cable penetration tube may have the following effects. Since small-medium nuclear reactors do not have sufficient space for structures supporting cables, sealing mechanisms configured to ensure the structural integrity of cables and at the same time to prevent leakage of cooling materials by means of penetration tube design have been required. According to the sealing mechanism for a small-medium RV cable penetration tube of the embodiments, the thimble 140 having a precise structure is placed around the cable 130 inserted into the RV 110 through the penetration tube 120, thereby preventing the cable 130 from being shaken and damaged by fluid flows and other internal vibrations occurring in the RV 110. Sealing mechanisms of the related art do not have devices for fixing cables. In general, power control cables used in nuclear reactors are relatively weak and vulnerable to fluid flows and other internal vibrations occurring in RVs. However, according to the sealing mechanism for a small-medium RV cable penetration tube of the embodiments, wobbling of the cable 130 may be prevented by the thimble 140 and the dimple groove portions 141 formed on the thimble 140. The thimble 140 are placed around the cable 130 in the penetration hole 121 and supported by the penetration tube end portion 122 and the other penetration tube end portion 124, and thus wobbling of the cable 130 may be prevented. Since the dimple groove portions 141 are in contact with the cable 130, the cable 130 may also be supported by the dimple groove portions 141, thereby improving the structural integrity of the cable 130. Since the structural integrity of the cable 130 is improved, the function of the cable 130 may not be damaged, and thus the functional integrity of the cable 130 may be improved. Various cables may be inserted into RVs, and such cables are required to be used for a long time without replacement so as to prevent leakage of reactor cooling materials. According to the embodiments, since the sealing mechanism for a small-medium RV cable penetration tube is configured to protect the cable 130, the cable 130 may be used for a long time without replacement, and thus maintenance and repair work may be performed within a shorter time period. In addition, owing to the simple structural design according to the embodiments, the cable 130 may be easily replaced with a new one. Thus, operators may work safely, and the modularity of facilities may be improved, thereby inducing the developing of relevant industries and increasing added value. In addition, when the cable 130 is repaired or replaced, the thimble 140 may function as a guide tube, and thus the repair or replacement of the cable 130 may be more easily performed. Small and medium nuclear reactors using small and medium reactor vessels are free from various problems associated with large reactors, and it is possible to construct such small and medium nuclear reactors within a relatively short period of time to stably supply electricity to necessary places. Thus, there is worldwide interest in small and medium nuclear reactors. However, it is difficult to install additional structures in small and medium nuclear reactors to support cables because of structural limitations. Thus, the integrity of cables may be guaranteed only by the method of providing precise structures in penetration tubes. The sealing mechanism for a small-medium RV cable penetration tube of the embodiments realizes the precise structural method by providing the thimble 140 and the dimple groove portions 141 in the penetration tube 120. That is, the sealing mechanism for a small-medium RV cable penetration tube of the embodiments improves the structural and functional integrity of the cable 130 based on the precise structural method. As described above, according to the one or more of the above embodiments, in the sealing mechanism for a small-medium RV cable penetration tube of the embodiments, the thimble 140 having a precise structure is placed on the cable 130 inserted into the RV 110 through the penetration tube 120, thereby preventing the cable 130 from being shaken and damaged by fluid flows and other internal vibrations occurring in the RV 110. In addition, the sealing mechanism improves the functional and structural integrity of the cable 130, and since the sealing mechanism has a simple structure, a time period necessary for maintenance and repair may be decreased. It should be understood that embodiments described herein should be considered in a descriptive sense only and not for purposes of limitation. Descriptions of features or aspects within each embodiment should typically be considered as available for other similar features or aspects in other embodiments. While one or more embodiments have been described with reference to the figures, it will be understood by those of ordinary skill in the art that various changes in form and details may be made therein without departing from the spirit and scope of the inventive concept as defined by the following claims.
060318937	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention is directed to a stray radiation grid, particularly for a medical X-ray apparatus of the type composed of a carrier material with absorption elements, particularly in the form of lead lamellae, that are arranged in rows spaced from one another and proceeding essentially parallel to one another, whereby the spacing between the rows of absorption elements is larger in the region of the edges of the grid than in the middle region. 2. Description of the Prior Art Stray radiation grids are employed in X-ray diagnostics for the suppression of stray radiation. The effectiveness of such a grid is particularly characterized by the line density (in lines per centimeter) and by its geometry, i.e. the ratio of height and thickness of the intermediate medium. This ratio is called the shaft ratio. In order to avoid a higher occlusion by the absorption elements, i.e., for example, the lead lamellae, in the outside regions than in the center, the grids are fashioned such that the absorption elements are aligned to the focus of the radiator, i.e. are "focussed". The focus spacing is thus a characteristic quantity of such grids. In these known grids, thus, the lead lamellae are arranged tilted. Alternatively, it is also known to conically erode the finished grid at one side, proceeding from the middle, and thus to modify the geometry. The known grids are composed of a carrier usually composed of paper; the absorption elements are usually lead lamellae. A disadvantage of the known embodiments is, for the first version described above, the manufacture thereof, since the lead lamellae arranged focussed, i.e. residing obliquely, must be brought into this focussing alignment in a complicated and extremely precise way. In the case of the slanted grid, the post-processing during manufacture is extremely involved. A stray radiation grid of the above-described type is described in U.S. Pat. No. 4,951,305. Given this grid, the spacing of the absorption elements of the respective grid or plane varies such that it is smaller in the middle of the grid than at the edge regions. A disadvantage, however, is that the grid exhibits a different absorption behavior over its area or surface caused by the increasing spacing of the absorption elements. SUMMARY OF THE INVENTION An object of the present invention is to provide a stray radiation grid with optimally uniform absorption behavior. This object is inventively achieved in a stray radiation grid of the type initially described, but wherein the width of the absorption elements is larger (thicker) in the region of the edges of the grid than in the middle. The grid thus departs from the extremely complicated tilting of the absorption elements, or beveling thereof, both of these techniques maintaining same spacing of the absorption elements. On the contrary, the absorption elements in the inventive stray radiation grid are seated closer to one another in the middle region than in the outer edge region, so that the shaft ratio is approximately balanced as a result, due to the oblique incidence of the image-active beam. A shaft ratio that is virtually constant over the entire grid is expediently achieved when the spacing from row to row increases continuously toward the edge proceeding from the middle of the grid. Since the absorption behavior changes with increasing spacing of the absorption elements from one another, it is inventively provided for compensation that the width of the absorption elements is larger in the region of the edges of the grid than in the middle, with the per element width increasing continuously toward the edges proceeding from the middle. On the basis of this measure, it is possible to realize a largely uniform absorption behavior over the entire grid width. The respective widths are thereby inventively selected such that they increase essentially proportionally to the increasing row spacing, i.e. such that the lead content per length unit remains constant over the entire grid width. In a further embodiment of the inventive grid, by contrast, the width increases sub-proportionately to the increasing row spacing. This inventive embodiment makes it possible to take the imaging radiation and the stray radiation which decrease toward the edge of an extensive grid (due to the distance square law), into consideration, so that a largely uniform absorption behavior is also established in the critical edge regions. Moreover, this embodiment allows the grid to be adapted to the decreasing dose rate in the beam cone, which decreases toward the edges. It is especially expedient, given an absorption property adapted to the actual conditions and given a constant setting of the shaft ratio, when the respective spacings between the absorption elements and/or the width of the absorption elements are inventively selected dependent on the local incident angle of the radiation, particularly X-ray radiation, in order to achieve a complete focussing with reference to the spacing from the radiation source. Compared to the continuous increase in spacing, in an alternative embodiment of the invention the grid, proceeding from the middle, has a number of regions within which the spacings between the rows of absorption elements are respectively constant, but the row spacing increases from region to region proceeding from the middle. This stray radiation grid is thus inventively constructed of separate segments that are respectively constant in terms of spacing; the spacing, however, the spacing increases from segment to segment. A precise focussing given substantial constancy of the shaft ratio can also be achieved with this inventive embodiment. Given this grid constructed of segments, the absorption elements are also thicker in the edge region than in middle. The width of the absorption elements can be essentially constant within a region, but can increase from region to region proceeding from the middle, as is also the case in the first embodiment of the invention. Here, too, there is the possibility for the width to increase essentially proportionately to the increasing spacing or, particularly given extensive grids, the width can increase subproportionately to the increasing spacing in order to adapt to the imaging radiation and the stray radiation decreasing at the edge side. Here, too, the spacing and/or width within the regions can be selected dependent on the incident angle of the radiation, particularly X-ray radiation, for a further improvement of the focussing.
	claims	1. A fuel bundle comprising:a first fuel rod, the first fuel rod includingfirst enriched uranium in a boost zone of the first fuel rod, the boost zone having a first average composition, the boost zone of the first fuel rod including a first burnable poison and constituting a bottommost portion of the first fuel rod, a percent of enrichment of the first enriched uranium in the boost zone of the first fuel rod being at least one percent,second enriched uranium in a second zone of the first fuel rod, the second zone having a second average composition, the second zone of the first fuel rod being arranged over the boost zone of the first fuel rod, the second average composition of the second zone being different from the first average composition of the boost zone, andnatural uranium in a third zone of the first fuel rod, the third zone having a third average composition, the third zone of the first fuel rod being arranged directly over the second zone of the first fuel rod, a length of the boost zone being equal to a length of the third zone, the third average composition of the third zone being different from the first average composition of the boost zone and the second average composition of the second zone. 2. The fuel bundle according to claim 1, further comprising:a second fuel rod, the second fuel rod includingthird enriched uranium in a boost zone of the second fuel rod, the boost zone of the second fuel rod being arranged directly at a bottom of the second fuel rod,fourth enriched uranium in a second zone of the second fuel rod, the second zone of the second fuel rod being arranged over the boost zone of the second fuel rod, andnatural uranium in a third zone of the second fuel rod, the third zone of the second fuel rod being arranged over the second zone of the second fuel rod, wherein a percent of enrichment of the third enriched uranium in the boost zone of the second fuel rod is equal to or greater than a percent of enrichment of the fourth enriched uranium in the second zone of the second fuel rod, and at least one of the boost and second zones of the second fuel rod includes a second burnable poison. 3. The fuel bundle according to claim 2, wherein the second burnable poison is gadolinia. 4. The fuel bundle according to claim 2, wherein the third zones of the first and second fuel rods have a length of about six inches. 5. The fuel bundle according to claim 2, wherein the second zones of the first and second fuel rods are directly on the boost zones of the first and second fuel rods. 6. The fuel bundle according to claim 2, wherein the boost zone of the second fuel rod does not include the second burnable poison. 7. The fuel bundle according to claim 2, wherein the boost zone and the second zone of the second fuel rod include the second burnable poison. 8. The fuel bundle according to claim 7, wherein a percent by weight of the second burnable poison in the boost zone is less than a percent by weight of the second burnable poison in the second zone. 9. The fuel bundle according to claim 2, wherein the first and second fuel rods are configured to yield a batch fraction of about 48% in a reactor core. 10. The fuel bundle according to claim 1, wherein the percent of enrichment of the first enriched uranium in the boost zone of the first fuel rod is about one percent. 11. The fuel bundle according to claim 1, wherein the percent of enrichment of the first enriched uranium in the boost zone of the first fuel rod is less than a percent of enrichment of the second enriched uranium in the second zone of the first fuel rod. 12. The fuel bundle according to claim 1, wherein the second zone has an enrichment that is greater than that of the boost zone. 13. The fuel bundle according to claim 1, wherein the length of the boost zone is about six inches. 14. The fuel bundle according to claim 1, wherein the first burnable poison is gadolinia.
	abstract	Refueling of a nuclear reactor (40) includes removing a fuel assembly (10). The removal method includes lowering a lifting tool (80) of a crane (44) onto a top of the fuel assembly. The lowered lifting tool including a plurality of downwardly extending elements (82) that surround and vertically overlap a portion (74) of a control rod assembly (70) extending above the top of the fuel assembly. The downwardly extending elements are locked with corresponding mating features (26) at the top of the fuel assembly to connect the lifting tool with the fuel assembly. The connected fuel assembly is moved into a spent fuel pool (42) using the crane, and the lifting tool is disconnected from the top of the fuel assembly by unlocking the downwardly extending elements from the corresponding mating features at the top of the fuel assembly.
	abstract	A compressor assembly and the method of using the same which includes an elongated spiral passageway within which a compact toroid plasma, such as a compact toroid plasma structure, can be efficiently compressed to a high-energy state by compressing the compact toroid plasma structure by its own momentum against the wall of the spiral passageway in a manner to induce heating by conservation of energy. The compressor assembly also includes a burn chamber that is in communication with the spiral passageway and into which the compressed compact toroid plasma structure is introduced following its compression.
	summary
047769823	claims	1. A method for temporary storage of radioactive material, said method comprising: transferring the radioactive material from a transport vehicle to a transfer container; moving said transfer container with the radioactive material to a storage room; positioning a double-walled storage container in the storage room; transferring the radioactive material directly from said transfer container to the positioned double-walled storage container; cooling the radioactive material in the double-walled storage container by circulating a heat exchange medium through the space formed between the double walls of said double-walled storage container to conduct heat away from the material; utilizing the heated heat exchange medium to provide an energy source. transferring the radioactive material from a transport vehicle to a transfer container; moving said transfer container with the radioactive material to a storage room; providing a double-walled storage container in the storage room; transferring the radioactive material directly from said transfer container to the doubled-walled storage container; storing the double-walled storage container containing the radioactive material in the storage room; cooling the radioactive material in the double-walled storage container by circulating a heat exchange medium through the space formed between the double walls of said double-walled container to conduct heat away from the material; utilizing the heated heat exchange medium to provide an energy source; and transferring the radioactive material from the double-walled storage container to another storage container after the radioactive material has cooled in the double-walled storage container. transferring the radioactive material from a transport vehicle to a transfer container; arranging a plurality of double-walled storage containers at predetermined locations in a storage room; moving said transfer container with the radioactive material to the storage room and adjacent one of the arranged double-walled storage containers; transferring the radioactive material directly from said transfer container to the adjacent one of the double-walled storage containers; storing the double-walled storage container containing the radioactive material in the storage room; cooling the radioactive material in the double-walled storage container by circulating a heat exchange medium through the space formed between the double walls of said double-walled container to conduct heat away from the material; utilizing the heated heat exchange medium to provide an energy source. 2. A method as in claim 1 including the step of stacking a plurality of said double-walled storage containers in said storage room. 3. A method as in claim 1 wherein the radioactive material is transferred from the transfer container to the double-walled storage container through a valve. 4. A method for temporary storage of radioactive material, said method comprising: 5. A method for temporary storage of radioactive material, said method comprising:
050323474	description	DESCRIPTION OF PREFERRED EMBODIMENT FIG. 1 shows a fuel assembly 1 comprising a framework consisting of spacer grids 2 spaced in the longitudinal direction of the assembly, guide tubes 3, to which the grids 2 are rigidly fixed, an upper joining piece 4 and a lower joining piece 5 which are fixed to the end of the guide tubes 3. The fuel rods 6 of the assembly which are shorter than the guide tubes 3 are positioned in the framework so as to form a bundle in which the rods are disposed parallel to one another. The rods are held laterally by the spacer grids 2 so as to form a uniform network of square mesh in the transverse sections of the assembly. The end grids 2a and 2b of the assembly differ in structure from the intermediate grids 2, as will be explained hereinbelow. In storage position or in operating position, the assembly is disposed vertically as shown in FIG. 1 and rests on a support by means of its lower joining piece 5. FIG. 2 shows an intermediate spacer grid 2 or mixing grid of the assembly shown in FIG. 1. Such a mixing grid consists of an assembly of small metal plates 8 disposed and assembled at right angles so as to form a network of square mesh, each of the cells 9 of which can receive a fuel rod 6 in order to position it in the network and hold it laterally. Some cells 7 are intended to receive a guide tube 3 which is fixed rigidly to the walls of the cell 7. The spacer grids 2 and the guide tubes 3 thus form a rigid framework which is capable of receiving the fuel rods 6. According to its outer contour whose square form corresponds to the section of the fuel assembly, the spacer grid 2 is delimited by a frame consisting of small plates 10 assembled at right angles according to the angles of the spacer grid 2. As may be seen in FIGS. 2 and 3A, the upper edge of each of the small plates 10 is cut in order to form successive fins 12 which are folded towards the inside of the spacer grid in order to form, with the plane of the small plate 10, an angle of perfectly defined value, which may be seen in FIG. 3A. In the case of a spacer grid of an assembly of developed structure, as shown in FIGS. 2, 3A and 4A, sixteen fins 12 are disposed on the upper edge of each of the small plates 10 of the square-section outer frame of the spacer grid, these fins 12 forming guide fins for the assembly during handling. As may be seen in FIG. 1, the lower edge of each of the small plates of the frame of a spacer 2 of the assembly comprises guide fins 13 which are also folded towards the inside of the spacer along the lower edge of the corresponding small plate in order to form an angle which is perfectly defined with the plane of this small plate 10 of the frame of the spacer grid. Each of the intermediate spacer grids 2 of the assembly also comprises, in the extension of the upper edge of each of the cells 9 receiving a fuel rods, fins 14, called mixing fins, which mix the cooling water circulating in contact with the rod where it emerges from the spacer grid. A homogenization of temperature of the water removing the heat provided by the fuel rods of the assemblies is thus obtained. Each of the cells 9 intended to receive a fuel rod comprises bosses 15 projecting towards the inside of the cell 9 and springs 16 generally comprising two active parts in two adjacent cells 9 of the spacer grid. The fuel rods 6 are held inside the cells 9 between the springs 16 and the bosses 15. In the case of a cell 7 intended to receive a guide tube 3, a sleeve 17 permits the rigid fixing, generally by crimping, of the guide tube on the spacer grid. In the case of the fuel assembly of advanced structure, such as that shown in FIGS. 2, 3A and 4A, the springs 16 are made of a nickel alloy which has great elasticity and very good behavioral characteristics in the environment of the nuclear reactor. These springs 16 are attached and fixed onto the small plates in the zones forming certain walls of the cells. The small plates made of zirconium alloy are assembled together to form the cells and stamped so as to enable the bosses 15 to be formed. In the case of an end spacer grid, such as 2a or 2b shown in FIG. 4A, the walls of the cells 9 intended to receive fuel rods do not comprise, in the extension of their upper edge, mixing fins 14. On the other hand, the small plates 10 forming the outer frame of these spacer grids comprise guide fins 12 and 13 inclined towards the inside of the spacer grid, according to a perfectly defined angle, in the same manner as the intermediate mixing grids 2. FIGS. 3B and 4B show two cells 9' adjacent to a plate 10' of the outer frame of a spacer grid having a conventional structure and different from the developed structure of the grid shown in FIGS. 2, 3A and 4A. Such a spacer grid of conventional structure consists solely of small plates made of zirconium alloy assembled in order to form a network of square mesh and in which the bosses 15' and the springs 16' for holding the fuel pencils are produced by stamping and shaping. In the case of intermediate mixing grids, such as shown in FIG. 3B, the upper parts of the small plates forming the walls of the cells are extended by mixing fins 14' which are inclined towards the inside of the cells 9'. On the other hand, in the case of spacer grids, such as shown in FIG. 4B, the upper parts of the small plates delimiting the cells do not comprise such mixing fins. In the case of intermediate spacer plates as well as end spacer plates, the small plates 10' forming the outer frame of the spacer grid comprise, on their upper edge as well as on their lower edge, guide fins 12' and 13' which are folded towards the inside of the spacer grid, forming an angle which is perfectly defined relative to the plane of the corresponding small plate 10'. However, in this type of conventional assembly, the fins 12' and 13' are not folded along an edge of the small plate 10' located at the level of the upper edges of the small plates delimiting the cells, but along a part of the small plate 10' located above this upper part of the cells. This construction, which is less satisfactory than that shown in FIGS. 3A and 4A as regards the resistance to deformation of the spacer plate, makes it possible to fold the guide fins 12' or 13' completely back against the inner surface of the corresponding small plate 10'. Quite obviously, this possibility does not exist in the case of the guide fins 12 and 13 of an assembly of developed structure, insofar as these fins 12 and 13 are located at the level of a small plate delimiting a row of cells of the grid. In this case, the guide fin 12 or 13 can be folded back towards the inside only by an angle of 90.degree.. This angle is generally in the region of 70.degree.. This thus results in difficulty in inserting the fuel rods and in circulating the cooling fluid at the entrance or at the exit of the spacer grid. In all cases, when the guide fins 12 and 13 or 12' and 13' are folded towards the inside, these fins no longer fulfil their guide function during handling of the assembly. The device according to the invention shown generally in FIG. 5 makes it possible to straighten the guide fins of the spacer grids of a fuel assembly located in a pool, when these guide fins have been deformed or folded back during handling of the assembly. The operation performed by the device according to the invention makes it possible to replace the guide fins in a position in which their inclination corresponds to the perfectly defined inclination according to the design of the spacer grid. FIG. 5 shows the device according to the invention in a position which makes it possible to straighten the guide fins of the spacer grids of a fuel assembly 1 located in a vertical position in a fuel assembly storage pool 20. The lower part of the fuel assembly 1 is engaged in a support device 21 which itself rests on a base 22. The device 21 has a certain conicity and is flared towards the top in order to facilitate insertion of the lower part of the fuel assembly 1. The work tool which makes it possible to straighten the guide fins of the spacer grids of the assembly 1, denoted generally by the reference 24, may be positioned at the level of the spacer grid 2 on which straightening is being performed by a device for support and displacement comprising a rod 25 or column of great length which is substantially vertical, along which a crossed carriage displacement assembly 26 carrying the work tool 24 can be displaced in the vertical direction. The crossed carriage device 26 also carries at least one video camera 27 which makes it possible to provide an image of the zone in which the fins of the spacer grid are being straightened. The upper part of the fuel assembly is placed under a depth of water of the order of four meters and the operations of positioning the work tool and straightening the fins are commanded from a control station 28 disposed above the upper level of the pool 20 and consisting of a platform on which are fixed the ends of the remote control devices 30 which make it possible to remotely activate the means for displacing the work tool 24 and the means for controlling this tool. The remote control means 30 consist of sheaths of great length in which it is possible to displace a maneuvering element, which is rigid in the pushing direction and deformable in flexion, guided by balls from the control station 28. Such a device, which is generally known as a ball remote control, is well known in the general field of remote controls. An operator 31 has handles which are accessible from the platform of the control station 28 in order to displace the crossed carriage assembly 26 in the vertical direction along the rod 25 in order to place the work tool 24 at the level of a spacer grid 2 on which the guide fins are being straightened. Moreover, the operator 31 can accurately position the work tool by remotely controlling the crossed carriages 26 whose displacements are guided in two directions which are perpendicular to one another and perpendicular to the substantially vertical axis of the rod 25. From the control station 28, the operator 31 can also command the activation of the tool for straightening the fins. The assembly of the device for positioning the work tool comprising the rod 25 and the crossed displacement carriages 26 is the subject of a patent application by the companies FRAMATOME and COGEMA filed on the same day as the present patent application. FIGS. 6, 7 and 8 show the crossed carriage device 26, on the upper part of which are fixed two brackets 36 and 37 in vertical position. The tool 35 for straightening the fins is supported by the brackets 36 and 37 by means of a bracket 38 and a guide support 39 which are fixed on the bracket 37 and on the bracket 36, respectively. The end of the sheath 40 of the ball remote control 30 which activates the straightening tool is fixed on the bracket 38 by means of two nuts 41. The guide support 39 delimits, in its inner part, a slide in which is mounted a sliding block 42 carrying a straightening hook 43 at its end. The slide of the guide support 39 has a direction which is substantially perpendicular to the axis of the rod 25. The hook 43 projects at the end of the slide of the support 39, this end of the support 39 being integrally attached to a bearing stop 45. The sliding block 42 comprises, at its upper end, a longitudinal groove 46 in which is engaged the end of a screw 47 passing through the upper part of the support 39. The sliding block 42 and the end hook 43 of this slide can be displaced in the direction of the slide perpendicular to the shaft 44 while being guided by the screw 47 engaged in the groove 46. The orientation of the sliding block and of the hook 43 about the axis of translation of the sliding block is thus fixed in this way. The element 48 forming the inner movable part of the ball remote control 30 is fixed, at its end, to a threaded rod 49 which is itself engaged in a threaded hole of a joining piece 50. The joining piece 50 comprises a T-shaped groove 51 in which an end part of corresponding form of the sliding block 42, opposite to the hook 43, engages. A screw 52 makes it possible to rigidly join the sliding block 42 and the joining piece 50. The bearing stop 45 is fixed on the support 39, at the end of the slide located towards the outside relative to the shaft 44 of the rod, by means of a support 54 extending the slide at its lower part and two screws 55 fixed in the support 39. FIGS. 6 and 9 show that, during use of the tool 35 for straightening a guide fin 12 of a spacer grid of the assembly, the stop 45 bears on the outer face of the small plate 10 of the frame of the spacer grid so that its upper edge is located exactly on the folding line of the guide fin 12. The hook 43 comprises a body 56 which is fixed by means of a screw 57 to the end of the sliding block 42. In FIG. 9, the fin 12 has been shown in solid lines in a position corresponding to its normal position whose inclination is perfectly defined, and in dotted lines in its deformed before straightening. The position of the end of the nose of the hook 43 relative to the bearing face of the stop 45, in the longitudinal direction of the sliding block, is adjusted so that the distance 58 between the inner end of the nose of the hook 43 and the bearing face of the stop 45 is greater than the distance separating the end of the fin 12 in its deformed position from the folding line of the small plate 10. On the other hand, the slope of the inner surface of the nose of the hook 43 corresponds to the inclination of the guide fin 12 in its correct position shown in solid lines and obtained after straightening, bearing in mind the elastic deformation of this fin. The hook 43 is engaged on the fin 12 to be straightened from above, the lower part of the bearing surface of the stop 45 coming into contact with the outer surface of the small plate 10. The assembly of the device for positioning the tool is then lowered in order to position the hook 43 in its position shown in FIG. 9. In this position, the end edge of the fin 12 is located slightly below the upper edge of the inner surface of the hook 43. As may be seen, in particular, in FIG. 10, the front part of the hook 43 comprises a flat 59 and two curved parts 60 so that the end of the hook can be inserted between the two fuel rods 6 disposed in the cells located on either side of the fin 12 to be straightened. The fin 12 is straightened by displacing the hook 43 by a perfectly determined amount, in the direction of the small plate 10. After this perfectly defined displacement, the distance 61 separating the end of the fin from the plate 10 corresponds to the perfectly defined correct position of this fin which is held in a straightened position by means of the inner surface of the hook 43 whose slope corresponds to the defined inclination of the fin 12. The sliding block 43 of the small plate 10 is displaced by pulling on the movable element 48 of the ball remote control 30, resulting in a displacement of the sliding block and of the hook 43 by a perfectly defined amount. The amount of displacement of the ball remote control is selected so as to very slightly exceed the defined position of the fin 12 which returns to its correct position, as a result of elasticity, after the force exerted by the hook 43, by means of the sliding block 42, has been released. FIG. 11 shows the section of the joining part 63 of the hook 43 between the body 56 and the nose delimited by the flats 59, as well as the hole 64 for fixing the hook 43 on the end of the sliding block 42 by means of screws 57. FIG. 12 shows a hook 65 comprising a nose 66 whose inclined inner surface 67 has an inclination which is much greater relative to the vertical than the inclined surface of the hook 43 shown in FIG. 7A. A hook such as 65 may be used to begin straightening of the fins when the latter are very inclined, final straightening being ensured by the nose of the hook 43 used after a first straightening of the fin using the hook 65. The form of the nose 66 of the hook may also be adapted in order to raise a fin when the latter is in a position inclined towards the inside of the spacer grid. FIG. 13 shows a work tool whose end which replaces the hook 43 of the device shown in FIGS. 6, 7 and 8 consists of a thrust device 68 comprising an end nose 69 having an inclined surface 70 which makes it possible to cause a fin 12 in a defective raised position, shown in dotted lines, to be changed to a position of correct inclination, shown in solid lines. In this case, the ball remote control is no longer activated in the pulling direction, but in the pushing direction after the end of the guide support 39 consisting of the stop 45 has been applied against the outer surface of the small plate 10. In this case, the guide fin 12 is pushed back towards the inside of the spacer grid until it reaches its position of correct inclination. This position is determined very precisely by adjusting the distance 71 separating the end of the fin from the inner face of the small plate 10. It is also possible to imagine a multipurpose tool which makes it possible simultaneously to push back fins which are excessively raised and to straighten fins which are folded towards the inside of the spacer grid. In this case, use will be made of a device combining straightening hooks, such as the hook 43, and a thrust device, such as the thrust device 68 shown in FIG. 13. The straightening hooks are engaged on the fins adjacent to the fin being straightened and a thrust device, placed in an intermediate position between the straightening hooks, is applied to the fin located in a central position. The various operations which have been described are checked by video cameras, such as the camera 27, carried by the crossed carriage device 26. An image of the zone in which the fins are being straightened is provided for the operator 31 by virtue of television screens. The operations performed on the fins of a fuel assembly are determined by inspecting the assembly inside the pool. Generally, the inclination defect of the fins is of the order of 10.degree. to 20.degree. and they can be straightened by using a hook 43, the inner surface of the nose of which has the desired inclination. In fact, the inner angle of the surface of the nose of the hook 43 will be slightly more acute in order to take into account the elasticity of the fin when the pulling force is released, as explained hereinabove. The amount of displacement of the movable part of the ball remote control of the tool is preadjusted so as directly to obtain the size adjustment corresponding to a correct straightening of the fin. The rapid opening obtained by virtue of the handle of the remote control makes it possible to release the hook from the fin, the support of the crossed carriage device 26 then being maneuvered vertically along the rod 25 in order to displace this device. Through out the operation for straightening the fin and during the operation for releasing the hook, the crossed carriage device is in a fixed position. The fins can also be straightened by using the device with a floating carriage. To this end, the tool for straightening the fins is mounted on the crossed carriage support, in its open position, i.e. with the straightening hook in its position separated from the end stop of the slide support. The hook is equipped with a miniaturized sensor which makes it possible to indicate to the operator by an indicator light located at the control station 28, that the end of the hook is correctly positioned on the fin 12. The straightening operation consists in maneuvering the crossed carriages so as to engage the hook on the fin and in lowering the support until the indicator light indicates a correct position. The carriage supporting the straightening tool is then disengaged in order to render it floating. Acting on the handle of the ball remote control 30 makes it possible to bring the bearing stop closer, by reaction, to the peripheral small plate 10 and the bearing hook closer to the guide fin. The fin is thus straightened by self-centering. In some cases, the fins may be greatly folded towards the inside of the spacer grid and, for example, in the case of a spacer grid of developed structure, shown in FIGS. 2, 3A and 4A, folded horizontally against the walls of the corresponding cells. In the case of an assembly of conventional structure, such as shown in FIGS. 3B and 4B, the guide fins 12' and 13' may even be folded towards the inside of the spacer grid beyond the horizontal position. In this case, use is made of a hook, such as shown in FIG. 12, which makes it possible to commence straightening of the fin in a first phase. In a second phase, the fin is straightened into its correct position by using a straightening hook of appropriate form. The device for positioning the work tool may be designed so as to receive, on its horizontal plate, two crossed carriage devices. In this case, one of the carriages can receive a tool comprising a hook with great inclination, which makes it possible to commence straightening of the fins, and the other carriage can receive a finishing tool which makes it possible to straighten the fin into its correct and well-defined inclination. In this case, work times are reduced and the phases of changing tools, which can be highly irradiated, are eliminated, thus reducing the time that the operators are exposed to ionizing radiation. The work tool can also comprise several hooks, this tool then having the form of a comb which is capable of acting simultaneously on several fins or even on the assembly of guide fins associated with a small plate of the outer frame of a spacer grid. To this end, the tool can comprise two to sixteen hooks disposed adjacent to one another. The tool which has been described can make it possible, by simply returning the guide support and the sliding block, to straighten fins located on the lower edge of the peripheral small plates of the spacer grid. If the device for positioning the tool comprises two crossed carriage devices, it is possible to fit different tools for straightening the lower fins and the upper fins, respectively. It is also possible to fit a tool, which makes it possible to straighten the fins, on one of the crossed carriage devices and a tool intended to push back the fins on the other crossed carriage. In this case, the tool for straightening the fins may be placed so as to provide a bearing reaction during the operation for pushing back the fins. Three remotely oriented cameras monitor all the operations for straightening or pushing back the fins, in order to follow the progression of these operations and, in particular, to monitor whether the fin being straightened is likely to break along the folding line. If an incipient breakage is observed, the fin will be removed by using, for example, a tool for recovering foreign bodies in the fuel assemblies comprising tongs, such as described in French Patent Application 88-09025. This device comprises a crossed carriage support which can be used to obtain the complete breakage of the fin. To this end, the end of the fin is gripped between the two parts of the tongs and the front carriage of the support device is displaced in an alternating to-and-fro movement in order to produce successive folding of the fin in either direction along the folding line showing the incipient breakage. This breaks the fin completely and the latter is deposited in a removal container placed in the vicinity of the assembly. It may be noted that the removal of some fins is unlikely to impair the normal operation of the reactor nor to substantially influence the safety of this reactor. The operation for overhauling the fins of the assembly can thus consist either in straightening a fin whose inclination is defective, if this fin does not show an incipient breakage, in particular, along its folding line, or in removing this fin if the latter is likely to break during the straightening operation. The invention is applicable to the overhauling of guide fins of any fuel assembly comprising spacer grids whose peripheral small plates comprise guide fins on their upper edge and/or on their lower edge.
059862769	claims	1. A system for eliminating X-ray radiation from power distribution tower equipment to ground, said system comprising: means for monitoring power distribution equipment-generated corona, said monitoring means comprising sensing instrumentation comprising an X-ray energy detector, means for transmitting the status data to a control center for accumulation and analysis, and means for correcting faults that are indicated by the analysis of the data, whereby X-rays produced by tower equipment faults are eliminated by prompt monitoring, reporting and correction of the fault, thus to prevent any damaging X-ray development. said sensing instrumentation further comprises an RF spectrum analyzer. said sensing instrumentation further comprises a corona detector including a wideband receiver and spectrum analyzer for detecting corona spectra. said sensing instrumentation further comprises an audio meter for detecting arcing sounds. monitoring the presence of corona generated in the distribution equipment from a control center, and applying low atomic number target materials to elevated-potential equipment for isolating potential X-ray targets to form safe discharge points for corona, whereby electrons accelerated by high voltage potentials have a high probability of interacting with the low Z material rather than penetrating to high Z material. said step for applying low atomic number target materials for isolating the potential X-ray generating target is a low Z hydrocarbon based plastic applied by at least one of the steps from the group comprising coating, spraying, dipping, extruding, chemical grafting, and wrapping. the applying of low atomic number target materials comprises the applying of a low Z material covering potential electrical discharge points by at least one of the group comprising extrusion, spraying, flame spraying, dipping, plating, sheathing as with woven graphite fibers, chemical grafting, and wrapping. said step for isolating potential X-ray targets by X-ray formation-inhibiting material comprises a low-X material covering applied to potential electrical discharge points by at least one step of the group of steps comprising extrusion, spraying, flame spraying, dipping, plating, sheathing as with woven graphite fibers, chemical grafting, and wrapping. shaping an electrostatic field of electrical potential by rearrangement of the electrical equipment to reduce peak field strength at potential X-ray targets, and applying an X-ray inhibiting material comprising low atomic number atoms onto said potential X-ray targets in the electrical equipment, whereby kinetic energy of the high speed electrons is converted into heat and low energy photons. the X-ray inhibiting material comprises polyarylene ether benzimidazole. means for shaping a field of the electrical potential of the electrical equipment to reduce peak field strength of potential X-ray targets, and means for isolating the potential X-ray targets by X-ray formation-inhibiting material comprising low atomic number atoms in a material placed between a source of the high speed electrons and the potential X-ray targets in the electrical equipment. said X-ray formation-inhibiting material comprises polyarylene ether benzimidazole. said means for isolating the potential X-ray target is a substrate configured to enclose a volume radiating the X-rays, and at least one low atomic number element suspended in said substrate. said means for isolating the potential X-ray target is an X-ray-safe target material for preventing X-ray radiation from electrical equipment, said target material comprising a substrate such as polymer polyvinyl acetate plastic gum adhesive for conforming and adhering to electrical equipment. said means for isolating the potential X-ray target is a low Z putty coating to prevent acceleration of high speed electrons to produce x-rays, while providing a soft target, said low Z putty comprising a polymer such as polyvinyl acetate mixed with powdered carbon. said means for isolating the potential X-ray target is a hydrocarbon based plastic material covering for electrical conductors. said means for isolating the potential X-ray target comprises low Z material contacts in switches forming the initial and breaking contacts of a switch to reduce X-ray generation present upon opening or closing a switch. said contacts are in parallel with conventional contacts to reduce contact arcing, and the low Z material contacts provide an initial striking contact and a final breaking contact with material that generates less X-rays when arcing than the high Z material switch contacts. 2. A system according to claim 1, wherein: 3. A system according to claim 2, wherein: 4. A system according to claim 1, wherein: 5. A method for eliminating radiation of X-rays from power distribution tower equipment to ground, comprising the steps of: 6. The method according to claim 5, wherein: 7. A method according to claim 5, wherein: 8. The method according to claim 5, wherein: 9. A method for eliminating radiation of X-rays by high speed electrons in high voltage fields from electrical equipment, comprising the steps of: 10. A method according to claim 9, wherein: 11. Apparatus for eliminating radiation of X-rays by high speed electrons in high voltage fields from electrical equipment, said apparatus comprising: 12. Apparatus according to claim 11, wherein: 13. Apparatus according to claim 11, wherein: 14. Apparatus according to claim 11, wherein: 15. Apparatus according to claim 11, wherein: 16. Apparatus according to claim 11, wherein: 17. Apparatus according to claim 11, wherein: 18. Apparatus according to claim 17, wherein:
	description	This application is a continuation of U.S. application Ser. No. 14/720,894, filed May 25, 2015, which claims the benefit of U.S. Provisional Application No. 62/002,922, filed May 26, 2014, each of which is hereby incorporated herein by reference in its entirety. The teachings provided herein are generally directed to systems and methods for obtaining nuclear fusion energy using a high energy charged particle convergence at a target cathode to increase the amount of fusion energy produced in a single fusion cycle. Most will agree that our world needs better sources of energy, source that are more efficient and would reduce the threat to the environment created by our current energy sources. In fact, most will agree that an uncompromising new energy architecture/paradigm is required to allow continued societal development and to avoid habitat and species loss. Current energy usage rewards a small minority of the population to the disadvantage of the majority and environmental quality. The combustion of carbon based fuels (coal, oil, natural gas) is still used primarily worldwide and still produces deleterious environmental effects in the form of elevated CO2 concentrations that is polluting our world and causing at least atmospheric warming and ocean chemistry changes. Ultimately, a future should be planned that addresses these issues as opposed to continued reliance on a strict capitalistic theory that will inevitably fail to meet the range of societal and environmental needs. An inexhaustible energy source will provide a basis for an economic structure that can be controlled without short-term tradeoffs that can be politically instituted. Many consider the possibility of the nuclear fusion power plant to be the best answer to the problem. One reason is that nuclear fusion is theoretically more efficient, requiring only about one millionth of the mass of fuel needed to produce the same amount of energy as a coal operating power plant. Another reason is that the fuel sources for nuclear fusion would be virtually unlimited, as these fuels are readily available. Another reason that nuclear fusion is desirable is that it doesn't suffer from diseconomies of scale—water and wind energy, for example, suffer from diseconomies as the optimal locations are used up and only less optimal locations remain, in addition to the fact that wind and water sources can vary, whereas fusion reactant sources remain reliable, as they are continuous, consistent, and abundant. Finally, it is believed that the nuclear fusion would offer a much safer process. For at least these reasons, the goal of producing fusion power to produce electricity has been pursued for decades and has been met with many problems that have not been solved; for example, there is still no controlled fusion process that can produce a sustained series of fusion reactions. FIGS. 1A and 1B illustrate nuclear fusion between deuterium and tritium, according to some prior art embodiments. As shown in FIG. 1A, the fusion between a first reactant that is deuterium, 2H, with a second reactant that is tritium, 3H, creates helium-4, 4He. The fusion also frees a neutron and releases 17.59 MeV of energy as heat. FIG. 1B shows the electrostatic force between the positively charged, cationic reactants, deuterium and tritium. FIGS. 2A and 2B illustrate state-of-the-art nuclear fusion reactors, according to some prior art embodiments. FIG. 2A illustrates a laser inertial fusion energy (LIFE) system 200 in which fusion 205 takes place in an evacuated 12 meter diameter steel chamber 210. Sixteen times a second, a 2 millimeter diameter target of deuterium-tritium fuel 215 is injected into the chamber, each target containing only about 0.7 mg of tritium, and each day about 1.3 million targets can be injected into the LIFE system 200 at a velocity of about 250 meters per second. And, the LIFE system 200 delivers 2 megajoule (MJ), 351 nanometer laser 230 pulses to indirect-drive fusion targets 215. The repeated fusion 205 reactions heat a lithium blanket surrounding the chamber, and the heat 220 generated, typically at about 600° C., is used to drive a steam turbine generator (not shown) to produce up to 1500 megawatts of baseload electricity from each plant. The chamber is filled with xenon gas 225 to protect the chamber 210 from ions and x-rays that are generated by the fusion 205 process in addition to the helium-4, 4He, and heat energy. Unfortunately, this technology has not been successfully scaled to produce a power plant, as the system is limited to use of a low reactant density of the deuterium and tritium gases which produces only random collisions and, thus, a low production of energy. Moreover, an unreasonably high energy is required to initiate the fusion, the system can only be cycled at a slow cycle frequency due to target loading and laser charging limitations, and there is no practical heat exchange method. It is a costly and inefficient system that leaves costly reactants unreacted due to the low reaction densities inherent to the design. FIG. 2B illustrates an international thermonuclear experimental reactor (ITER) system 250 in which fusion takes place in a donut-shaped vacuum vessel called a tokamak 260. An electromagnet (not shown) conducts electricity through the center of the tokamak 260 to produce a voltage (not shown) across gas reactants deuterium, 2H, and tritium, 3H, that have been injected 265 in the tokamak 260, the voltage ripping electrons from the deuterium, 2H, and tritium, 3H, to ionize the deuterium, 2H, and tritium, 3H, into cationic reactants. The ionized, cationic reactants form a plasma 270. Magnetic coils 275 are used to compress and confine the plasma 270 to keep it away from the walls of the tokamak 260 and protect the tokamak 260 from the high temperatures that are developed in the plasma. The magnetic coils 275 also generate a current in the plasma 270, heating it to 10 million° C. which, unfortunately, is still not hot enough for fusion to occur. To raise the temperature to 100-200 million° C., which is hot enough for fusion to occur, radiowaves and microwaves 280 are fired into the plasma. Unfortunately, this technology has not been successfully scaled to produce a power plant, and the system is also limited to the use of deuterium and tritium gases at a low reactant density which produces a low reaction rate and low production of energy. In addition, an especially problematic condition is that the reaction products are not removed but, rather, mix with the reactants and slow the reaction rate. Moreover, cycling of the process is impractical due to the large volume of the system which demands a long pump-out time. As with the LIFE system 200, there is no practical heat exchange method in the ITER system 250, and the ITER system 250 is costly to manufacture, maintain, and operate due to its complexity and inefficiencies. Although there are theoretical designs for a reactor that is hoped to deliver ten times more fusion energy than the amount needed to heat up the plasma 270 to the required temperatures for fusion to occur, the ITER facility is still not expected to finish its construction phase until at least 2019 and is not expected to begin full deuterium-tritium fusion until at least 2027. Given the above, it should be appreciated that those of skill will appreciate a controlled fusion process that can produce a sustained series of fusion reactions. Namely, a process that (i) uses a substantially higher reactant density of the deuterium and tritium gases by converging cationic reactants into the higher reaction density at a target cathode rather than relying on random collisions, the converging producing a substantially higher rate of fusion and energy production; (ii) uses a substantially lower input of energy to initiate the fusion; (iii) can be cycled at a substantially higher cycle frequency; (iv) has a practical heat exchange method; (v) is substantially less costly to manufacture, operate, and maintain; and, (vi) has a substantially improved reaction efficiency as a result of not mixing reactants with products. The teachings provided herein are generally directed to systems and methods for obtaining nuclear fusion energy using a high energy charged particle convergence at a target cathode to increase the amount of fusion energy produced in a single fusion cycle. Namely, the teachings provide a controlled fusion process that can produce a sustained series of fusion reactions: a process that (i) uses a substantially higher reactant density of the deuterium and tritium gases by converging cationic reactants into the higher reaction density at a target cathode rather than relying on random collisions, the converging producing a substantially higher rate of fusion and energy production; (ii) uses a substantially lower input of energy to initiate the fusion; (iii) can be cycled at a substantially higher cycle frequency; (iv) has a practical heat exchange method; (v) is substantially less costly to manufacture, operate, and maintain; and, (vi) has a substantially improved reaction efficiency as a result of not mixing reactants with products. For example, the teachings include a method of producing an at least substantially continuous electrical energy from a cyclized nuclear fusion reaction, comprising evacuating a reaction chamber to a pressure that is lower than about 10−3 torr; inducing a pulse of (i) a first reactant into the evacuated reaction chamber through a first reactant port and a pulse of (ii) a second reactant into the evacuated reaction chamber through a second reactant port; and, converging the first reactant with the second reactant at a target cathode for colliding and fusing the first reactant with the second reactant to create a heat energy. The converging can include, for example, creating an electrical field in the reaction chamber by applying a voltage across an anode surface positioned in the interior of the reaction chamber and a cathode surface positioned in the interior of the reaction chamber, the electric field ionizing the first reactant to generate a cationic first reactant and ionizing the second reactant to generate a cationic second reactant. In addition, the converging can include establishing a negative charge on the target cathode for attracting and converging the cationic first reactant and the cationic second reactant at the target cathode for colliding and fusing the cationic first reactant with the cationic second reactant to create the heat energy. The method can include transferring the heat energy to a steam vessel to drive a turbine to create an electrical energy. The method can be cyclic by replacing the target cathode with a replacement target cathode to complete a first cycle of the nuclear fusion method; and, repeating the evacuating, inducing, applying, converging, transferring, and replacing for n additional cycles of the nuclear fusion method, wherein n is an integer that produces an at least substantially continuous electrical energy from the nuclear fusion reaction. One of skill will appreciate that the first reactant and second reactant can be any reactant useful in producing a fusion reaction using the methods and systems taught herein. For example, the first reactant and second reactant can each be independently selected from the group consisting of deuterium, tritium, and helium-3, boron-11, lithium-6, and a proton, in some embodiments. In some embodiments, the first reactant and the second reactant are independently selected from the group consisting of deuterium, tritium, and helium. In some embodiments, the first reactant is deuterium and the second reactant is tritium. In some embodiments, the first reactant is deuterium and the second reactant is deuterium. In some embodiments, the first reactant is tritium and the second reactant is tritium. In some embodiments, the first reactant is deuterium and the second reactant is helium-3. In some embodiments, the first reactant is helium-3 and the second reactant is helium-3. In some embodiments, the first reactant is a proton and the second reactant is boron-11. And, in some embodiments, the first reactant is a proton and the second reactant is lithium-6. One of skill will appreciate that the pressure in the reaction chamber can be varied to any pressure that one of skill will find useful in the methods and systems provided herein. For example, the pressure in the evacuated reaction chamber can range from about 10−4 torr to about 10−9 torr in some embodiments, and from about 10−6 torr to about 10−9 torr in some embodiments. The teachings are also directed to a system that can be used in practicing the methods taught herein. For example, the teachings include a system for producing an at least substantially continuous electrical energy from a cyclized nuclear fusion reaction. In some embodiments, the system comprises a reaction vessel having a reaction chamber configured for evacuation of the chamber to a pressure that is lower than about 10−3 torr; a vacuum port adapted for an operable connection to a vacuum source for evacuating the reaction chamber to a pressure that is lower than about 10−3 torr; a first injector in operable communication with a first reactant port in the evacuated reaction chamber for inducing a pulse of a first reactant into the evacuated reaction chamber through the first reactant port; a second injector in operable communication with a second reactant port in the evacuated reaction chamber for inducing a pulse of a second reactant into the evacuated reaction chamber through the second reactant port; an anode surface and a cathode surface for operably connecting to a voltage source, the anode surface and the cathode surface positioned in the interior of the reaction chamber to create an electric field in the evacuated reaction chamber upon application of a voltage, the electric field ionizing the first reactant to generate a cationic first reactant and ionizing the second reactant to generate a cationic second reactant; a target cathode positioned in the reaction chamber at a first distance from the first injector and a second distance from the second injector, the target cathode configured to function as a negatively charged electrode for attracting and converging the cationic first reactant and the cationic second reactant at the target cathode for colliding and fusing the cationic first reactant with the cationic second reactant to create a heat energy; a steam chamber in operable contact with the reaction chamber, the steam chamber configured for receiving the heat energy from the fusion reaction in the reaction chamber. One of skill will also appreciate that the first reactant port and the second reactant port can include a configured nozzle, designed for a particular embodiment. For example, nozzle flow design can be varied to change the shape and speed of the first reactant from the first reactant port and the shape and speed of the second reactant from the second reactant port. In some embodiments, the pulse of the first reactant or the pulse of the second reactant is applied as a convergent flow on the target electrode. In some embodiments, the pulse of the first reactant or the pulse of the second reactant is applied as a divergent flow on the target electrode. And, in some embodiments, the pulse of the first reactant or the pulse of the second reactant is applied as a fan pattern on the target electrode. It should also be appreciated that the nozzle design can be independently selected for each of the first injector and the second injector. For example, the first injector can be configured for injecting deuterium and the second injector can be configured for injecting tritium. The first injector can be configured for injecting deuterium and the second injector can be configured for injecting deuterium. The first injector can be configured for injecting tritium and the second injector can be configured for injecting tritium. Likewise, one of skill will appreciate that the reaction chamber can be configured to operate at any pressure that one of skill will find useful in the methods and systems provided herein. For example, the reaction chamber can be configured to operate in the pressure range from about 10−4 torr to about 10−9 torr in some embodiments, and from about 10−6 torr to about 10−9 torr in some embodiments. One of skill will appreciate that the positioning of the first reactant port, the second reactant port, and the target cathode can be adjusted to vary the first distance between the first reactant port and the target cathode and the second distance between the second reactant port and the target electrode. In some embodiments, the first distance and the second distance are at least substantially the same. In some embodiments, the first distance and the second distance are varied to calibrate and synchronize the collision between the cationic first reactant and the cationic second reactant. Likewise, one of skill will also appreciate that the negative charge on the target cathode can be varied to calibrate and synchronize the collision between the cationic first reactant and the cationic second reactant. Moreover, one of skill can vary the first distance, the second distance, and the charge on the target cathode to calibrate and synchronize the collision between the cationic first reactant and the cationic second reactant. One of skill will appreciate that the size of the target cathode can be varied for any of a variety of operational considerations. In some embodiments, for example, the target electrode can have an area ranging from about from about 1.00×10−10 m2 to about 1.00×10−6 m2. One of skill will appreciate that the target cathode can be constructed of a variety of different materials. For example, the target cathode can be any conducting material. In some embodiments, the target cathode can comprise a metal. In some embodiments, for example, the target cathode can be comprised of aluminum, an aluminum alloy, or copper. In some embodiments, the target cathode can be comprised of a metal selected from the group consisting of aluminum, antimony, barium, bismuth, boron, carbon (e.g., amorphous, diamond, graphene, graphite), cadmium, calcium, chromium, cobalt, copper, gold, iridium, iron, lead, magnesium, manganese, mercury, molybdenum, nickel, platinum, potassium, rhenium, silver, sodium, steel, tantalum, tellurium, tin, titanium, tungsten, uranium, vanadium, zinc, and alloys thereof. The target cathode can comprise a semiconductor or conductive polymer, in some embodiments. In some embodiments, one or more shields can be used to protect the first reactant port and the second reactant port from the heat and/or products of the fusion reaction. As such, in some embodiments, the systems can further comprise a shield between the first injector and the target electrode, between the second injector and the target electrode, or a combination thereof. Systems and methods are provided herein for obtaining nuclear fusion energy using a high energy charged particle convergence at a target cathode to increase the amount of fusion energy produced in a single fusion cycle. Namely, a controlled fusion process is provided that can produce a sustained series of fusion reactions: a process that (i) uses a substantially higher reactant density of the deuterium and tritium gases by converging cationic reactants into the higher reaction density at a target cathode rather than relying on random collisions, the converging producing a substantially higher rate of fusion and energy production; (ii) uses a substantially lower input of energy to initiate the fusion; (iii) can be cycled at a substantially higher cycle frequency; (iv) has a practical heat exchange method; (v) is substantially less costly to manufacture, operate, and maintain; and, (vi) has a substantially improved reaction efficiency as a result of not mixing reactants with products. FIGS. 3A and 3B illustrate a system for obtaining nuclear fusion energy using a high energy charged particle collision, according to some embodiments. In some embodiments, the systems and methods can produce an at least substantially continuous electrical energy from a cyclized nuclear fusion reaction. As shown in FIG. 3A, the system can comprise a reaction vessel 300 having a reaction chamber 305 configured for evacuation of the chamber 305 to a pressure that is lower than about 10−3 torr; a vacuum port 310 adapted for an operable connection to a vacuum source 315 for evacuating the reaction chamber 305 to a pressure that is lower than about 10−3 torr; a first injector 320 in operable communication with a first reactant port 325 in the evacuated reaction chamber 305 for inducing a pulse of a first reactant, tritium (3H) 329 into the evacuated reaction chamber 305 through the first reactant port 325; a second injector 340 in operable communication with a second reactant port 345 in the evacuated reaction chamber 305 for inducing a pulse of a second reactant 349, deuterium (2H) into the evacuated reaction chamber 305 through the second reactant port 345; an anode surface 360 and a cathode surface 365 for operably connecting to a voltage source (not shown), the anode surface 360 and the cathode surface 365 positioned in the interior of the reaction chamber 305 to create an electric field in the evacuated reaction chamber 305 upon application of a voltage 360,365, the electric field ionizing the first reactant 329 to generate a cationic first reactant 330 and ionizing the second reactant 349 to generate a cationic second reactant 350; a target cathode 365 positioned in the reaction chamber 305 at a first distance (distance between 325 and 365) from the first injector 325 and a second distance (distance between 345 and 365) from the second injector 340, the target cathode 365 configured to function as a negatively charged electrode for attracting and converging the cationic first reactant 330 and the cationic second reactant 350 at the target cathode 365 for colliding and fusing the cationic first reactant 330 with the cationic second reactant 350 to create a heat energy; a steam chamber 380 in operable contact with the reaction chamber 305, the steam chamber 380 configured for receiving the heat energy 375 from the fusion 355 reaction in the reaction chamber 305. The voltage applied to attract and converge the cationic reactants 330,350 at the target cathode 365 can be referred to as the “bias”, in some embodiments, indicating the use of voltage to generate a force to transport the cationic reactants 330,350 to the target cathode 365 for the collision and fusion. FIG. 3B provides an enlarged view of the target cathode 365 and it's insulator 367. The cationic first reactant 329, tritium (3H) and the cationic second reactant 329, deuterium (2H) are shown as attracted and converging to the target cathode 365 for colliding and fusing to create the heat energy 375. As opposed to the random distribution and random collision present in current state of the art fusion technologies, the systems and methods provided herein substantially increase the reactant density, and fusion rate, when compared to the current state of the art fusion technologies. In some embodiments, each of the reactant density and fusion rate can be said that it is “substantially greater than” a state-of-the-art process, or can be referred to as “substantially increasing” over a state-of-the-art process, when it increases by about 2×, about 3×, about 5×, about 10×, about 15×, about 20×, about 25×, about 30×, about 40×, about 50×, about 60×, about 70×, about 80×, about 90×, about 100×, about 200×, about 300×, about 400×, about 500×, about 600×, about 700×, about 800×, about 900×, or any amount therein in increments of 1×. Likewise, in some embodiments, each of the reactant density and fusion rate can be said that it is “substantially greater than” a state-of-the-art process, or can be referred to as “substantially increasing” over a state-of-the-art process, when it increases by about 103, about 104, about 105, about 106, about 107, about 108, about 109, about 1010, about 1011, about 1012, about 1013, about 1014, about 1015, or any amount therein in increments of 0.500×103. The reactant density and fusion rate can be measured and compared to a state-of-the-art process in any manner known to one of skill to provide an acceptable comparable measure. For example, the exothermicity of the fusion output, the energy produced, normalized to the amount of reactants injected, can provide an acceptable comparable measure of at least fusion rate, and possibly an indirect comparable measure of reactant density, in some embodiments. Likewise the electricity produced from a steam turbine that uses the energy from the fusion to produce electrical power per, based on a normalization of the amount of reactants injected, can provide an acceptable comparable measure of at least fusion rate, and possibly an indirect comparable measure of reactant density, in some embodiments. The efficiency of the process can also be considered in the comparable measures, for example, perhaps to further normalize the comparisons to further compare the instant systems and methods to the current state-of-the-art using reaction efficiency, cost efficiency, and the like. One of skill can use any metric that is considered acceptable in the art to compare the instant systems and methods to the current state-of-the-art, for example, the ITER system or the LIFE system. The small size and negative charge of the target focuses the convergence of the reactants. It should be appreciated that the surface area of the target cathode can be varied as a process variable in order to adjust, for example the reaction density, rate of reaction, and/or energy produced by the fusion reaction. One of skill will appreciate that any area that works with the principles of the teachings provided herein can be used. In some embodiments, the surface area of the target can be defined as the front and back surface only, disregarding the surface on the edge of the target. In some embodiments, the target cathode can have any configuration that serves to attract the first reactant and the second reactant in a convergent manner to at least one point of collision. In some embodiments, there is more than one point of collision and, in some embodiments, there are many points, areas, or planes of collision such as more than 3, more than 5, more than 10, more than 20, more than 50, more than 100, and so on. The number of points of collision can be a variable selected to increase the operation efficiency of the system. For example, in some embodiments, the target cathode can be in the form of a scaffolding, cage, or mesh structure each carrying a negative charge on areas that represent a variety of planes for collision, as opposed to a planar structure with, perhaps, two primary planar surfaces carrying a negative charge, such as the target cathode disc shown in FIG. 3B. The target cathode disc shown in FIG. 3B has a front planar surface, a back planar surface, and a small surface around the circumference of the disc. As the charged particles approach the disc from different directions, attracted by the negative charge, the angles of their approaches to impinge on the charged surface of the disc are highly variable. A basket, screen, or mesh target cathode may facilitate the impingement of the cationic reactants from just about any angle of impingement on the target cathode that may occur in the reaction chamber. Moreover, reactants can come close to the target cathode and miss it, and a cage-type structure, or a structure having even more charged surfaces in layers, may facilitate a higher efficiency of impingement of the cationic reactants on the target cathode. In some embodiments, a cage can be any shape desired, such as a spherical cage, ellipsoid cage, cubical cage, polyhedral cage, conical cage, cylindrical cage and the like. The cage can be used as a second component in combination with a primary target cathode, for example, in combination with the disc-shaped cathode in FIG. 3B. The primary cathode in FIG. 3B, for example, can have a cylindrical or conical cathode cage added to both sides of the primary target cathode to help capture, confine or focus the beams of reactants approaching the primary target cathode. Such a secondary component can carry a positive charge to help focus the cationic reactants toward the cathode, in some embodiments. In some embodiments, the area of the target can be on the surface of a disc, a sphere, an ellipsoid, cube, polyhedron, and the like, as well as the same or similar shapes but manufactured using screen or mesh materials to provide several conductive surfaces in the form of a scaffolding, cage, or basket, for example, that can be approached from about any angle of impingement that may occur from a reactant. In some embodiments, the surface area can range from about 1.3648×10−10 m2 to about 1.3648×10−6 m2, from about 1.3648×10−9 m2 to about 1.3648×10−7 m2, from about 1.00×10−10 m2 to about 1.00×10−6 m2, about 1.3648×10−8 m2, about 1.00×10−8 m2, about 10−11 m2, 10−10 m2, 10−9 m2, 10−8 m2, 10−7 m2, 10−6 m2, 10−5 m2, or any range therein. One of skill will appreciate that the voltage between an anode surface and a cathode surface in the systems and methods can be varied and are selected to be large enough to (i) create a sufficient electric field to ionize the reactants, (ii) overcome the like-charge repulsion to enable the first reactant and the second reactant to collide sufficiently for fusion, and (iii) drive the convergence of the reactants to create the high reactant density at the target cathode and measured as the average density over the entire surface area of the target cathode. As such, in some embodiments, where the voltage is between the target cathode and an anode surface and is sufficiently large enough to create the convergence of the reactants towards the target cathode and the force of collision required for the fusion between the first reactant and the second reactant. Likewise, the voltage between the target cathode and the anode surface can also be used in the creation of the electric field, in some embodiments, to ionize the first reactant and the second reactant into cationic first reactant and cationic second reactant, respectively. In some embodiments, the voltage can range from about 10 kV to about 30 MV, from about 15 kV to about 30 MV, from about 20 kV to about 30 MV, from about 30 kV to about 30 MV, from about 15 kV to about 25 MV, from about 15 kV to about 20 MV, from about 10 kV to about 10 MV, from about 15 kV to about 10 MV, from about 15 kV to about 5 MV, from about 40 kV to about 5 MV, from about 50 kV to about 5 MV, from about 100 kV to about 5 MV, from about 250 kV to about 5 MV, from about 500 kV to about 5 MV, from about 500 kV to about 2 MV, from about 1 MV to about 5 MV, from about 1 MV to about 3 MV, from about 1 MV to about 2 MV, or any range of voltages therein in increments of 1 kV. In some embodiments, the voltage can be about 10 kV, about 11 kV, about 12 kV, about 13 kV, about 14 kV, about 15 kV, about 20 kV, about 25 kV, about 30 kV, about 35 kV, about 40 kV, about 45 kV, about 50 kV, about 60 kV, about 70 kV, about 80 kV, about 90 kV, about 100 kV, about 200 kV, about 300 kV, about 400 kV, about 500 kV, about 600 kV, about 700 kV, about 800 kV, about 900 kV, about 1 MV, about 2 MV, about 3 MV, about 4 MV, about 5 MV, about 10 MV, about 15 MV, about 20 MV, about 25 MV, about 30 MV, or any voltage therein, or range of voltages therein, in increments of 1 kV. In some embodiments, the voltage is greater than about 15 kV, greater than about 20 kV, greater than about 25 kV, greater than about 30 kV, greater than about 35 kV, greater than about 40 kV, greater than about 45 kV, greater than about 50 kV, or greater than any kV between 15 kV and 50 kV in increments of 1 kV. The kV can be constant, or it can be varied, in the operation of a system or method taught herein. Variable frequency electric fields can also be used. In some embodiments, a variable frequency field can be used, for example, to increase the ionization efficiency of the ionization step to create the cationic reactants. Given the teachings of the systems and methods provided herein, it should also be appreciated that the design is adapted to provide a substantially higher reactant density than currently provided by the state-of-the-art. And, one of skill will appreciate that, as the reactant density increases, the energy output of the system will substantially increase per cycle over the current state-of-the-art processes, and the total energy output of the system will likewise substantially increase over the current state-of-the-art processes. The substantially higher performance of the systems and methods taught herein over the current state-of-the-art processes can be established in any manner considered acceptable to one skilled in the art. In some embodiments, for example, the reactant density can represent mass/volume and can range from about 1.5×103 g/cm3 to about 1.5×1010 g/cm3 normalized as an average reactant density over the entire surface of the target electrode. In some embodiments, the reactant density can range from about 1.5×105 g/cm3 to about 1.5×1010 g/cm3 normalized as an average reactant density over the entire surface of the target electrode. In some embodiments, the reactant density can range from about 1.5×108 g/cm3 to about 1.5×1010 g/cm3 normalized as an average reactant density over the entire surface of the target electrode. In some embodiments, the reactant density can range from about 1.5466×1010 g/cm3 normalized as an average reactant density over the entire surface of the target electrode. This is a very significant, surprising and unexpected increase in reactant density over the current state-of-the-art, as the LIFE system (See FIG. 2A) has a reactant density of only about 1000 g/cm3, according to some measures. In some embodiments, the substantially higher reactant density of the instant systems and methods as compared to the current state-of-the-art processes can be represented by using a measure of the monolayer of nuclei that converge on the surface of the target cathode. In some embodiments, for example, the monolayer reactant density can range from about 103 nuclei/m2 to about 1029 nuclei/m2 normalized as an average monolayer reactant density over the entire surface of the target electrode. In some embodiments, for example, the monolayer reactant density can range from about 104 nuclei/m2 to about 1029 nuclei/m2 normalized as an average monolayer reactant density over the entire surface of the target electrode. In some embodiments, for example, the monolayer reactant density can range from about 105 nuclei/m2 to about 1028 nuclei/m2 normalized as an average monolayer reactant density over the entire surface of the target electrode, or any range therein. In some embodiments, for example, the monolayer reactant density can be about 103 nuclei/m2, 104 nuclei/m2, 105 nuclei/m2, 106 nuclei/m2, 107 nuclei/m2, 108 nuclei/m2, 109 nuclei/m2, 1010 nuclei/m2, 1011 nuclei/m2, 1012 nuclei/m2, 1013 nuclei/m2, 1014 nuclei/m2, 1015 nuclei/m2, 1016 nuclei/m2, 1017 nuclei/m2, 1018 nuclei/m2, 1019 nuclei/m2, 1020 nuclei/m2, 1021 nuclei/m2, 1022 nuclei/m2, 1023 nuclei/m2, 1024 nuclei/m2, 1025 nuclei/m2, 1026 nuclei/m2, 1027 nuclei/m2, 1028 nuclei/m2, 1029 nuclei/m2, or any range within these values, each monolayer reactant density normalized as an average monolayer reactant density over the entire surface of the target electrode. In some embodiments, for example, the monolayer reactant density be about 5.2×1028 nuclei/m2 normalized as an average monolayer reactant density over the entire surface of the target electrode. One of skill will appreciate that the reactor should be made of a material that takes into consideration the high temperatures and pressures present from the nuclear fusion reaction. Any material that meets this criteria can be used. For example, the reaction chamber can be made of steel. Moreover, the stresses in the system can be reduced by operating under steady state conditions where possible to avoid inducing unnecessary thermal stresses in the materials. It should also be appreciated that most any component of the systems and methods taught herein can be subject to his criteria, and in particular those materials that form a part of the reaction chamber. Methods of using such systems are also provided herein. The methods can comprise, for example, evacuating the reaction chamber 305 to a pressure that is lower than about 10−3 torr; inducing a pulse of (i) the first reactant 329 into the evacuated reaction chamber 305 through a first reactant port 325 and a pulse of (ii) a second reactant 348 into the evacuated reaction chamber 305 through the second reactant port 345; and, converging the first reactant 329 with the second reactant 349 at the target cathode 365 for colliding and fusing 355 the first reactant 329 with the second reactant 349 to create the heat energy 375. The converging can include, for example, creating an electrical field in the reaction chamber by applying a voltage (not shown) across the anode surface 360 positioned in the interior of the reaction chamber 305 and the cathode surface 365 positioned in the interior of the reaction chamber 305, the electric field ionizing the first reactant 329 to generate the cationic first reactant 330 and ionizing the second reactant 349 to generate the cationic second reactant 350. In addition, the converging can include establishing a negative charge on the target cathode 365 for attracting and converging the cationic first reactant 330 and the cationic second reactant 350 at the target cathode 365 for colliding and fusing 355 the cationic first reactant 330 with the cationic second reactant 350 to create the heat energy 375. The methods will generally include transferring the heat energy 375 to a steam vessel 380 to drive a turbine (not shown) to create an electrical energy. One of skill will appreciate that there are several variations possible in the implementation of these process steps in series. Table 1 is illustrative of some of the variations. TABLE 1StepEmbodiment 1Embodiment 2Embodiment 3Embodiment 41Valve OpenVoltageVoltageValve Openinducedinduced2Gas leavesValve OpenValve OpenGas leavesmanifoldmanifold3VoltageGas leavesGas leavesVoltageinducedmanifoldmanifoldinduced4Ionize gasIonize gasValve closesIonize gas5Gas acceleratesGas acceleratesIonize gasGas acceleratesto targetto targetto target6Gas hits targetGas hits targetGas acceleratesGas hits targetto target7Fusion occursFusion occursGas hits targetFusion occurs8Produce 4HeProduce 4HeFusion occursProduce 4Heand energyand energyand energy94He is4He isProduce 4HeValve closesevacuatedevacuatedand energy10Valve closesValve closes4He is4He isevacuatedevacuated11Manifolds fillManifolds fillManifolds fillManifolds fill One of skill will appreciate that the primary steps of the methods taught herein will often include (i) inducing a pulse of the first reactant, (ii) inducing a pulse of the second reactant; (iii) ionizing the first reactant and the second reactant; (iv) converging the first reactant and the second reactant on the target cathode; and (v) collecting heat energy from the fusion reaction. Processes of cycling the fusion reaction include the step of evacuating the 4He from the reaction chamber. The possible variations around these primary steps are, of course, numerous in many embodiments, and understood as mere process variations by those of skill. The method can be cyclic by replacing the target cathode with a replacement target cathode to complete a first cycle of the nuclear fusion method; and, repeating the evacuating, inducing, applying, converging, transferring, and replacing for n additional cycles of the nuclear fusion method, wherein n is an integer that produces an at least substantially continuous electrical energy from the nuclear fusion reaction. One of skill will appreciate that the number of cycles that can be run is a process variable that can depend on materials used to construct the reaction vessel and, thus the operational constraints of the reaction vessel, operational constraints of peripheral components, the preventative maintenance schedule set for the equipment, and the like. As such, assuming at least the fusion rate of the LIFE system which is 1,382,400 fusions/day (16 fusions/second) and a shutdown for repairs no more than once per quarter, n can be about 124,416,000 cycles for a single reactor. In some embodiments, n can range from about 10 to about 10,000,000,000 cycles for a single reactor. In some embodiments, n can range from about 100 to about 1,000,000,000 cycles for a single reactor. In some embodiments, n can range from about 1000 to about 100,000,000 cycles for a single reactor. In some embodiments, n can range from about 10,000 to about 10,000,000 cycles for a single reactor. In some embodiments, n can be about 10, 102, 103, 104, 105, 106, 107, 108, 109, 1010, 1011, or 1012 cycles, or any range of cycles therein, for a single reactor. The practice of the method includes selecting the first reactant and the second reactant. One of skill will appreciate that the first reactant and second reactant can be any reactant useful in producing a fusion reaction using the methods and systems taught herein. For example, the first reactant and second reactant can each be independently selected from the group consisting of deuterium, tritium, and helium-3, boron-11, lithium-6, and a proton, in some embodiments. In some embodiments, the first reactant and the second reactant are independently selected from the group consisting of deuterium, tritium, and helium. In some embodiments, the first reactant is deuterium and the second reactant is tritium. In some embodiments, the first reactant is deuterium and the second reactant is deuterium. In some embodiments, the first reactant is tritium and the second reactant is tritium. In some embodiments, the first reactant is deuterium and the second reactant is helium-3. In some embodiments, the first reactant is helium-3 and the second reactant is helium-3. In some embodiments, the first reactant is a proton and the second reactant is boron-11. And, in some embodiments, the first reactant is a proton and the second reactant is lithium-6. One of skill will appreciate that the pressure in the reaction chamber can be varied to any pressure that one of skill will find useful in the methods and systems provided herein. For example, the pressure in the evacuated reaction chamber can range from about 10−4 torr to about 10−9 torr in some embodiments, and from about 10−6 torr to about 10−9 torr in some embodiments. In some embodiments, the pressure in the evacuated reaction chamber can be about 10−3 torr, about 10−4 torr, about 10−5 torr, about 10−6 torr, about 10−7 torr, about 10−8 torr, about 10−9 torr, or any range therein. It should be appreciated that the high vacuum conditions existing in the reaction chamber permit the use of high purity reactants that result in an optimized fusion reaction. The term “purity” can be referred to as the absence of the reaction product helium-4 from the prior reaction remaining in the trajectory of the reactants entering the reaction chamber. The purity is obtained by evacuating the reaction chamber between fusion cycles. Likewise, the kinetic energy of the reactants traveling to the target cathode for collision is also optimized by the high vacuum condition, because the reactants are likewise allowed to accelerate while remaining unimpeded by the helium-4 product remaining in the reaction chamber from the prior cycle. Moreover, one of skill will understand that any appropriate vacuum system can be used. For example, such a system can be composed of conventional components. In some embodiments, turbomolecular or diffusion high vacuum pumps backed by rotary vane pumps may be used for the evacuation of the reaction chamber. A Roots-type blower pump may also be used, in some embodiments, to assist in the evacuation of the reaction chamber. In some embodiments, an appropriate vacuum system operation for transient conditions may include (i) a pumpdown with a venting to atmospheric pressure to eliminate a back-diffusion of vacuum pump lubricants to assure a contamination-free reaction chamber condition. In some embodiments the vacuum pump system can include three vacuum pumps in series, a high vacuum pump such as a turbomolecular or diffusion pump, backed by a Roots-type pump, backed by a rotary vane or piston-type pump. One of skill will appreciate that the positioning of the first reactant port and the second reactant port, relative to the placement of the target cathode, can be varied. For example, the distance from a reactant port to the target can be selected by considering various system parameters, such as duty cycle, duty cycle number, duty cycle frequency, and the like; power applied to the system, power applied for a voltage, power applied for the electric field, power applied for the converging, power output measured, power output desired, and the like; maintenance, number and frequency of maintenance cycles used, number and frequency of maintenance cycles desired, and the like; performance, performance measured, performance of power output, performance desired, and any performance specification in general. Analogous to a combustion engine, the timing of the relative release of the first reactant and second reactant can be adjust to “tune” the fusion “engine”. The relative amount of each injection, relative timing of each injection, relative pressure of injection, relative frequency of injection, the design of each injector, such as injection nozzle, and the like, are examples of parameters that can be varied and manipulated, along with the first distance and second distance, to optimize the energy output, and other performance parameters. Such tuning of the “fusion engine” (i.e. any system taught herein), for example, can be used to optimize system performance measured in terms of energy output, economy of operation, life of the fusion engine, frequency of maintenance required, profit from the operation, or any combination thereof, in some embodiments. In some embodiments, the positioning of each of the system components can be adjusted to vary the first distance between the first reactant port and the target cathode and the second distance between the second reactant port and the target electrode. In some embodiments, the first distance and the second distance are at least substantially the same. In some embodiments, the first distance and the second distance are varied to calibrate and synchronize the collision between the cationic first reactant and the cationic second reactant. In some embodiments, the first distance and the second distance can be independently selected to range from about 0.001 meter to about 30 meters, from about 0.01 meter to about 20 meters, from about 0.001 meter to about 10 meters, from about 0.1 meter to about 15 meters, from about 0.1 meter to about 12 meters, or any range therein in increments of 0.1 meter. In some embodiments the first distance and second distance can be independently selected to be about 0.001 meter, about 0.01 meter, about 0.1 meter, about 0.5 meter, about 1.0 meter, about 2.0 meters, about 3.0 meters, about 4.0 meters, about 5.0 meters, about 6.0 meters, about 7.0 meters, about 8.0 meters, about 9.0 meters, about 10.0 meters, about 11.0 meters, about 12.0 meters, about 13.0 meters, about 14.0 meters, about 15.0 meters, about 16.0 meters, about 17.0 meters, about 18.0 meters, about 19.0 meters, about 20.0 meters, or any distance therein in increments of 0.1 meter. One of skill will also appreciate that the first reactant port and the second reactant port can include a configured nozzle, designed for a particular embodiment. For example, nozzle flow design can be varied to change the shape and speed of the first reactant from the first reactant port and the shape and speed of the second reactant from the second reactant port. In some embodiments, the pulse of the first reactant or the pulse of the second reactant is applied as a convergent flow on the target electrode. In some embodiments, the pulse of the first reactant or the pulse of the second reactant is applied as a divergent flow on the target electrode. And, in some embodiments, the pulse of the first reactant or the pulse of the second reactant is applied as a fan pattern on the target electrode. The reactant injectors can be designed for the injection of a particular reactant, to vary the amount, speed, configuration, or direction of injection, and the like. Any parameter associated with an injector can be varied, including pressure of injection, amount of reactant feed to the injector, the dwell time of the injection, and the like. This adjustment of amount injected and dwell time of injection might be considered somewhat analogous to the fuel injection system and cam design of a combustion engine. For at least these reason, one of skill will appreciate that the nozzle design can be independently selected for each of the first injector and the second injector. For example, the first injector can be configured for injecting deuterium and the second injector can be configured for injecting tritium. Likewise, the first injector can be configured for injecting deuterium and the second injector can be configured for injecting deuterium. Moreover, the first injector can be configured for injecting tritium and the second injector can be configured for injecting tritium. In some embodiments, the opening and closing of the valve that feeds an injector can be referred to as a valve actuation cycle that includes opening the valve to a fully open position, maintaining the open position for a brief interval of time resulting in the steps of the first reactant and the second reactant entering the reaction chamber, the first reactant and the second reactant ionizing, the first reactant and the second reactant accelerating to the target cathode, the first reactant and the second reactant fusing to create fusion energy, and then the valve closing and staying closed until the start of the next cycle. The dwell time of a reactant feed through an injector, for example, the first injector or the second injector, is a variable that controls the time it takes to open the valve, how long the valve stays open, and the time it takes to close the valve. As such, the dwell time can be adjusted to control how much reactant enters the reactant chamber. In some embodiments, the dwell time to open a valve to feed an injector can range from about 0.01 millisecond to about 100 milliseconds, from about 0.1 millisecond to about 10 milliseconds, from about 1.0 millisecond to about 10 milliseconds, from about 0.1 millisecond to about 5 milliseconds, from about 0.01 millisecond to about 1.0 millisecond, or any range therein in increments of 0.01 millisecond. In some embodiments, the dwell time to maintain the open valve to feed an injector can range from about 0.01 millisecond to about 100 milliseconds, from about 0.1 millisecond to about 10 milliseconds, from about 1.0 millisecond to about 10 milliseconds, from about 0.1 millisecond to about 5 milliseconds, from about 0.01 millisecond to about 1.0 millisecond, or any range therein in increments of 0.01 millisecond. In some embodiments, the dwell time to close a valve that feeds an injector can range from about 0.01 millisecond to about 100 milliseconds, from about 0.1 millisecond to about 10 milliseconds, from about 1.0 millisecond to about 10 milliseconds, from about 0.1 millisecond to about 5 milliseconds, from about 0.01 millisecond to about 1.0 millisecond, or any range therein in increments of 0.01 millisecond. In some embodiments the dwell time can refer to a “total dwell time”, which is the sum of the time to open, time remaining open, and time to close the valve. As such, in some embodiments, the total dwell time can also refer to a range of about 0.03 milliseonds to about 300 milliseconds, from about 0.01 millisecond to about 100 milliseconds, from about 0.1 millisecond to about 10 milliseconds, from about 1.0 millisecond to about 10 milliseconds, from about 0.1 millisecond to about 5 milliseconds, from about 0.01 millisecond to about 1.0 millisecond, or any range therein in increments of 0.01 millisecond. Likewise, one of skill will also appreciate that the negative charge on the target cathode can likewise be varied, increasing the electron density on the cathode apart from the voltage, to have further calibration and synchronization control over the collision between the cationic first reactant and the cationic second reactant at the target cathode. Moreover, one of skill can vary the first distance, the second distance, as well as the charge on the target cathode to calibrate and synchronize the collision between the cationic first reactant and the cationic second reactant. One of skill will appreciate that this calibration and synchronization of collisions might be considered somewhat analogous to the timing the fuel input, ignition, and position of the piston in the combustion chamber to optimize the performance of a combustion engine. One of skill will appreciate that the target cathode can be constructed of a variety of different materials. For example, the target cathode can be any conducting material. In some embodiments, the target cathode can comprise a metal. In some embodiments, for example, the target cathode can be comprised of aluminum or an aluminum alloy. In some embodiments, the target cathode can be comprised of a metal selected from the group consisting of aluminum, antimony, barium, bismuth, boron, carbon (e.g., amorphous, diamond, graphene, graphite), cadmium, calcium, chromium, cobalt, copper, gold, iridium, iron, lead, magnesium, manganese, mercury, molybdenum, nickel, platinum, potassium, rhenium, silver, sodium, steel, tantalum, tellurium, tin, titanium, tungsten, uranium, vanadium, zinc, and alloys thereof. The target cathode can comprise a semiconductor or conductive polymer, in some embodiments. In some embodiments, the target can comprise water or a conductive plasma. In some embodiments, one or more shields can be used to protect the first reactant port and the second reactant port from the heat and/or products of the fusion reaction. As such, in some embodiments, the systems can further comprise a shield between the first injector and the target electrode, between the second injector and the target electrode, or a combination thereof. FIGS. 4A and 4B illustrate a reactor node having multiple reactant ports protected by a reactant nozzle shield in a system and method for obtaining nuclear fusion energy using a high energy charged particle collision, according to some embodiments. FIG. 4A illustrates an expanded view of a shield 405 that is used to protect multiple reactant ports, each of which are used for inducing pulses of reactants, such as deuterium, 2H, and tritium, 3H, into a reaction chamber (not shown). The shield 405 protects the multiple reactant ports 410 from the output of the fusion reaction to avoid any deleterious effects on the reactant ports including, for example, the deposition of neutrons or damage from direct exposure to the exothermic output from the fusion. FIG. 4B is a cross-sectional view of a reaction chamber 400 of the use of the shield 405 to protect multiple reactant ports inside the reaction chamber 400. As shown the pulses of reactants that include the deuterium, 2H, and tritium, 3H, which are injected into the reaction chamber 400 and ionized by an electric field to form a cationic deuterium, 2H, reactant and a cationic tritium, 3H, reactant. The shield 405 protects the multiple reactant ports 410 from the output of the fusion reaction to avoid the deleterious effects on the reactant ports. The electrode 415 carries a negative charge as the target cathode, and the cationic deuterium and cationic tritium converge to the target cathode 415 for collision and fusion. FIG. 4B can be referred to as a single reactor “node”, in some embodiments. In some embodiments, a “reactor node” can be defined as having (i) a target cathode; (ii) at least two reactant ports; (iii) a chamber wall for conducting the heat energy out of the reactor vessel; and (iv) an anode to establish the voltage with the target cathode, wherein the anode can be the chamber wall. In some embodiments, a “reactor node” can be defined as having (i) a target cathode; (ii) at least two reactant ports; (iii) a shield to protect the at least two reactant ports; (iv) a chamber wall for conducting the heat energy out of the reactor vessel; and (v) an anode to establish the voltage with the target cathode, wherein the anode can be the chamber wall. FIG. 5 is a process flow chart for obtaining nuclear fusion energy using a high energy charged particle collision through a reactor node having multiple reactant ports protected by a reactant nozzle shield, according to some embodiments. FIG. 5 is a representation of the reactor node of FIG. 4B, using the illustration of FIG. 4B to shown the process steps. In step 1, 505, the reactants enter the reaction chamber the deuterium, 2H, and tritium, 3H, which are injected into the reaction chamber 400 and are ionized step 2, 510, by an electric field to form a cationic deuterium, 2H, reactant and a cationic tritium, 3H, reactant. In some embodiments, photonic energy may also be applied to supplement the ionization of the reactants. The cationic deuterium, 2H, reactant and a cationic tritium, 3H, accelerate to the electrode which is the target cathode 415 in step 3, 515. The shield 405 protects the multiple reactant ports 410 from the output of the fusion reaction that occurs in step 4, 520, to avoid the deleterious effects on the reactant ports 410. The electrode 415 carries a negative charge as the target cathode, and the cationic deuterium and cationic tritium converge to the target cathode 415 for collision and fusion. Steps 1-4, 515-520, represent merely one cycle of fusion. In some embodiments, the fusion reactor has multiple cycles, and the cycles can be separated by an additional step of evacuating the reaction chamber to remove fusion products, such as 4He and neutrons, and increase the performance of the next reaction cycle. The neutrons may stick to the walls of the reaction vessel. A single reactor node can have a single target cathode, in some embodiments. However, in some embodiments, a single reactor node can have more than one target cathode. In some embodiments, for example, a single reactor node might have 2, 3, 4, 5, 6, 7, 8, 9, 10, or more target cathodes. In some embodiments, a single reactor node can have one or more cage or mesh type target cathodes, a configuration that may be implemented for at least the reasons taught herein. In some embodiments, a reactor vessel can have more than one reactor node. In some embodiments, the reactor vessel can have a single reaction chamber with more than a single node, for example, 2, 3, 4, 5, 6, 7, 8, 9, 10, or more reactor nodes. In some embodiments a reactor vessel can have a single reaction chamber or more than one reaction chamber. In some embodiments, for example, the reactor vessel might have 2, 3, 4, 5, 6, 7, 8, 9, 10, or more reaction chambers. Each reaction chamber in each reactor vessel can have multiple reactor nodes. FIG. 6 is a cross-sectional top view (or bottom view) of a multi-node, high vacuum reactor for obtaining nuclear fusion energy using a high energy charged particle collision, each reactor node having multiple reactant ports protected by a reactant nozzle shield and a target electrode, according to some embodiments. In FIG. 6, the multi-node, high vacuum reactor 600 has 6 reactor nodes, only one of which is labeled in the diagram for purposes of clarity, the others serving as merely additional nodes having the same structure for purposes of the illustration. It is possible, however, for each of the nodes to have an independent configuration that is not the same as all other nodes, at least in some embodiments. The system has a single reaction chamber 605 is shared by each of the 6 reactor nodes and configured for evacuation of the reaction chamber 605 to a pressure that is lower than about 10−3 torr; a vacuum port 610 adapted for an operable connection to a vacuum source (not shown for evacuating the reaction chamber 605 to a pressure that is lower than about 10−3 torr; a first injector (not shown) in operable communication with a first reactant port 625 for a first reactant (shown in dotted lines behind a shield 607) in the evacuated reaction chamber 605 for inducing a pulse of a first reactant (not shown) into the evacuated reaction chamber 605 through the first reactant port 625; a second injector (not shown) in operable communication with a second reactant port 645 for a second reactant (shown in dotted lines behind a shield 607) in the evacuated reaction chamber 605 for inducing a pulse of a second reactant (not shown) into the evacuated reaction chamber 605 through the second reactant port 645; an anode surface 660 and a cathode surface 665, which in this embodiment is the target cathode, for operably connecting to a voltage source (not shown), the anode surface 660 and the cathode surface 665 positioned in the interior of the reaction chamber 605 to create an electric field in the evacuated reaction chamber 605 upon application of a voltage 660,665, the electric field ionizing the first reactant (not shown) to generate a cationic first reactant (not shown) and ionizing the second reactant (not shown) to generate a cationic second reactant (not shown); a target cathode 665 positioned in the reaction chamber 605 at a first distance (distance between 625 and 665) from the first injector 625 and a second distance (distance between 645 and 665) from the second injector 640, the target cathode 665 configured to function as a negatively charged electrode for attracting and converging the cationic first reactant (not shown) and the cationic second reactant (not shown) at the target cathode 665 for colliding and fusing the cationic first reactant (not shown) with the cationic second reactant (not shown) to create a heat energy 675; a steam chamber 680 in operable contact with the reaction chamber 605, the steam chamber 680 configured for receiving the heat energy (not shown) from the fusion reaction (not shown) in the reaction chamber 605. The voltage applied to attract and converge the reactants at the target cathode 665 can be referred to as the “bias”, in some embodiments, indicating the use of voltage to generate a force to transport the cationic reactants (not shown) to the target cathode 665 for the collision and fusion. FIGS. 7A-7E include a perspective view of a portion of a multi-node, high vacuum reactor for obtaining nuclear fusion energy using a high energy charged particle collision, each reactor node having multiple reactant ports protected by a reactant nozzle shield and a target electrode, in addition to insulated target electrode assemblies, according to some embodiments. In FIG. 7, the multi-node, high vacuum reactor 600 of FIG. 6 is shown from the perspective view for a better view of the relationship between the target cathode 665 as it relates to its insulator 667, shield 607, the reactant ports 625,645, and the wall 606 of the reaction chamber 605 in each the 6 reactor nodes. FIGS. 7A-7C illustrate an insulator 667 for a conducting wire 663 that is operably connected to the target cathode 665. FIG. 7D illustrate a cross-section of the insulator 667 for the conducting wire 663 that is operably connected to the target cathode 665. FIG. 7E illustrates a side view of the distal end of the insulator 667 for the conducting wire 663 that is operably connected to the target cathode 665. All dimensions shown are not limiting, are for example only, and are in meters. FIG. 8 shows an expanded view of a portion of a multi-node, high vacuum reactor for obtaining nuclear fusion energy using a high energy charged particle collision, including reactor nodes in a high vacuum reactor chamber, reactant injector ports with injector valves for each node, a target tray with cassettes providing replacement target electrodes, and a steam chamber for capturing and moving nuclear fusion energy in the form of steam to a steam turbine to create electricity, according to some embodiments. The multi-node, high vacuum reactor 800 has 5 reactor nodes in the high vacuum reactor chamber 805 as can be seen from the illustration. The reactor chamber 805 is surrounded by a steam chamber 850 for the energy capture and transfer of energy to a steam turbine, for example, to produce electricity. A target load/switch-out assembly 890 with cassettes 895 carrying extra target cathodes 865 is also provided. Also included is a set of one type of reactant valves 899 which, in this case are designed as damped deuterium-tritium manifolds with reactant staging regions. It should be appreciated that any valve configuration known to one of skill can be used in the process of inducing a pulse of reactant into the reaction chamber 805. The following FIGS. 9A-13 provide more detail on the components of the multi-node, high vacuum reactor 800. All dimensions shown are not limiting, are for example only, and are in meters. FIGS. 9A-9F shows a multi-node, high vacuum reactor chamber having reactant injector ports for each node, and an insulated target electrode used in each node, according to some embodiments. FIG. 9A is an expanded view of the reactor chamber 805, including the first injector port 813 and second injector port 815. The reactor chamber 805 includes reinforcing ribs 823 to provide strength to the reactor chamber 805 under the extreme temperature and pressure from the fusion reaction. The reactor chamber 805 lid 809 also has the reinforcing ribs 823, as well as the insulator 867 for the conducting wire (not shown) that is operably connected to the target cathode (not shown). FIG. 9B is an end-view of the reactor chamber 805, FIG. 9C is a side-view of the reactor chamber 805, and FIG. 9D is a side-view of the lid 809. FIGS. 9E and 9F illustrate a perspective view and a side view of the distal end of the insulator 867 for the conducting wire 863 that is operably connected to the target cathode 865. All dimensions shown are not limiting, are for example only, and are in meters. FIGS. 10A-10D illustrate a target load/switch-out assembly with cassettes providing replacement target electrodes for 5 reactor nodes, according to some embodiments. FIG. 10A illustrates a perspective view of the target load/switch-out assembly 890. The target load/switch-out assembly 890 provides new target cathodes for each reaction cycle, replacing the target cathode consumed in the prior cycle. The replacement target cathodes can be supplied automatically at atmospheric conditions but are transferred to the reaction chamber nodes at high vacuum pressure levels. The target load/switch-out assembly contains the following components: a cassette loading chamber (qty 5; one for each node) 891, a first buffer chamber 892 (qty 5; one for each node), a target transfer chamber 893 (qty 5; one for each node), a second buffer chamber 894 (qty 5; one for each node), and a cassette unload chamber 895 (qty 5; one for each node). The progression 896 of changing the target cathode (not shown) follows the general path in the component series from the cassette loading chamber 891 (shown with door 891D open) to the buffer chamber 892 to the target transfer chamber 893 to the buffer chamber 894, and finally to the cassette unload chamber 895. The target load/switch-out assembly 890 also contains a target transfer actuator housing 897 containing an actuator 898 for each one of the reactor nodes 1-5. In FIG. 10A, the identifiers “Node 1”, “Node 2”, “Node 3”, “Node 4”, and “Node 5” identifier the tray progression 896 lanes for each of nodes 1-5. FIG. 10B shows an end-view of the target load/switch-out assembly 890. FIG. 100 shows a top-view of the target load/switch-out assembly 890. FIG. 10D shows a side view of the target load/switch-out assembly 890. All dimensions shown are not limiting, are for example only, and are in meters. FIGS. 11A-11D illustrate cassettes for the target load/switch-out assembly 890 for changing out the spent target cathodes from each fusion reaction, each cassette providing five (5) target trays, each of the 5 target trays providing sixteen (16) replacement target electrodes, according to some embodiments. FIG. 11A shows a target tray cassette 1100 containing five (5) target trays 1105, each target tray 1105 containing sixteen (16) target cathodes 1167. FIG. 11B shows an end-view of the target tray 1105. FIG. 11C shows a top-view of the target tray 1105. FIG. 11D shows a side-view of the target tray 1105. The target cassette 1100 can be held together using any operable means such as, for example, by a tray frame or scaffolding 1101. The cassette loading chamber 891 is configured to receive the target cassette 1100 through a door 891D. The cassette loading chamber 891 is also configured to be evacuated to intermediate vacuum levels (significantly below atmospheric but above 10−3 torr). After evacuation, the cassette loading chamber 891 then feeds target trays 1105 into the first buffer chamber 892 which is configured to provide additional outgassing of the target tray 1105 and it's replacement target cathodes 1167. The target tray 1105 moves from the first buffer chamber 892 into the target transfer chamber 893. The target tray 1105 is then indexed in x-y directions in the target transfer chamber 893 to align each of the 16 target cathodes 1167 with the actuator 898 which moves the respective replacement target cathode in the z-direction through the target transfer actuator housing 897 to transfer the respective replacement target cathode 1167 into the reactor chamber. In some embodiments, for example, the respective replacement target cathode 1167 is moved into the reactor chamber, and any remains of the spent target cathode can be retrieved. The target tray 1105 is then progressed into the second buffer chamber 894 and then the cassette unload chamber 895 which can function to accept an unloaded target cassette frame 1101 at atmospheric pressure, evacuate to intermediate vacuum levels, receive target trays 1105 into the target cassette frame 1101, potentially having remainders of spent targets from the second buffer chamber 894 for removal from the system. This target replacement process can be repeated for each cycle at each of the 5 nodes. It should be appreciated that the mechanism of target cathode transfer can be any mechanism, there can be any number of nodes, and that this mechanism is merely an example of the cyclic process of replacing spent target cathodes in the reaction chamber for a sustained series of fusion reactions. FIGS. 12A-12D illustrate a steam chamber for capturing and moving nuclear fusion energy in the form of steam to a steam turbine to create electricity, according to some embodiments. FIG. 12A shows an expanded view of the steam chamber 850 with lid 859. The steam chamber 850 is also designed to accommodate the five (5) reactor nodes of the high vacuum reactor chamber 805 as can be seen from the illustration. The reactor chamber 805 is surrounded by the steam chamber 850 for the energy capture and transfer of energy to a steam turbine, for example, to produce electricity. Water goes into the steam chamber 850 and leaves as steam out of the steam chamber 850, as illustrated. A target load/switch-out assembly 890 with cassettes 895 carrying extra target cathodes 865 is also provided. The steam chamber 850 must also be designed to accommodate the passage of the first injector port 813 and second injector port 815 into the reaction chamber 805 (only one side is marked in FIG. 12A for clarity, but the other side is complementary and can be assumed to be a mirror image of the marked side). Like the reactor chamber 805, the steam chamber 850 also includes reinforcing ribs 823 to provide strength to the reactor chamber 805 under the extreme temperature and pressure from the fusion reaction. The steam chamber lid 859 also has the reinforcing ribs 823, as well as the insulator 867 for the conducting wire (not shown) that is operably connected to the target cathode (not shown). FIG. 12B is an end-view of the steam chamber 850, FIG. 12C is a side-view of the steam chamber 850, and FIG. 12D is a side-view of the lid 859. All dimensions shown are not limiting, are for example only, and are in meters. FIG. 13 illustrates a cross-sectional bottom view of a multi-node, high vacuum reactor for obtaining nuclear fusion energy using a high energy charged particle collision, including reactor nodes in a high vacuum reactor chamber, reactant injector ports for each node with mass flow controllers, and a steam chamber for capturing and moving nuclear fusion energy in the form of steam to a steam turbine to create electricity, according to some embodiments. The five (5) node reactor can use conventional mass flow controllers 1333 to provide precise reactant flow into the reactant chamber 805. The mass flow controllers 1333 can be calibrated to complement the reactant valves 899 which, in this case are designed as the damped deuterium-tritium manifolds with reactant staging regions. The process gas flow is illustrated, with the deuterium, 2H, and the tritium, 3H, flowing through the mass flow controllers 1333 and into the reaction chamber 805 for the fusion reaction that produces the energy that converts the “water in” to the “steam out” as illustrated to drive a steam turbine (not shown) and generate electricity. Also illustrated is the evacuation of helium-4, 4H, from the reaction chamber 805. Without intending to be limited to any theory or mechanism of action, the following examples are provided to further illustrate the teachings presented herein. It should be appreciated that there are several variations contemplated within the skill in the art, and that the examples are not intended to be construed as providing limitations to the claims. The equation describing the energy balance for a representative 2 gigawatt (GW) power plant is as follows and assumes a 100% utilization of reactants: 17.6 ⁢ ⁢ MeV ⁢ ⁢ 2 H ⁢ ( single ⁢ ⁢ molecule ) + 3 H ⁢ ⁢ ( single ⁢ ⁢ molecule ) = 2.8198 × 10 - 18 ⁢ ⁢ megajoule ⁢ ⁢ 2 ⁢ ⁢ G ⁢ ⁢ Watt = ( 2.8198 × 10 - 18 ⁢ ⁢ megajoule ) * ( x ⁢ / ⁢ sec ) ⁡ [ 2 H ⁢ ( single ⁢ ⁢ molecule ) + 3 H ⁢ ( single ⁢ ⁢ molecule ) ] ⁢ ⁢ ( x ⁢ / ⁢ sec ) ⁡ [ 2 H ⁢ ( single ⁢ ⁢ molecule ) + 3 H ⁢ ( single ⁢ ⁢ molecule ) ] = 7.0926 × 10 20 ⁢ / ⁢ sec ⁢ ( x ⁢ / ⁢ sec ) ⁡ [ 2.0141 ⁢ ⁢ u ⁢ ⁢ 2 H + 3.0160 ⁢ ⁢ u ⁢ ⁢ 3 H ] = 7.0926 × 10 20 ⁢ / ⁢ sec ⁢ ( x ⁢ / ⁢ sec ) ⁡ [ 3.3234 × 10 - 24 ⁢ ⁢ gram ⁢ ⁢ 2 H + 5.0082 × 10 - 24 ⁢ ⁢ gram ⁢ ⁢ 3 ⁢ H ] = ( 2.3572 ⁢ ⁢ milligram ⁢ ⁢ 2 H + 3.5522 ⁢ ⁢ milligram ⁢ ⁢ 3 H ) ⁢ ⁢ / ⁢ sec = [ 2.3572 ⁢ ⁢ milligram ⁢ ⁢ 2 H ⁢ ( 22.4 ⁢ ⁢ liter ⁢ / ⁢ 2.0141 ⁢ ⁢ gram ⁢ ⁢ 2 H ⁢ / + 3.5522 ⁢ ⁢ milligram ⁢ ⁢ 3 H ⁢ ( 22.4 ⁢ ⁢ liter ⁢ / ⁢ 3.0160 ⁢ ⁢ gram ⁢ ⁢ 3 H ) ] ⁢ / ⁢ sec = ( .02622 ⁢ ⁢ liter ⁢ ⁢ 2 H + .02638 ⁢ ⁢ liter ⁢ ⁢ 3 H ) ⁢ / ⁢ sec ( ref ⁢ : ⁢ ⁢ 1 ) Accordingly, for a 2 GW power plant, the fuel flow rate should be approximately=1572.9 sccm2H+1582.9 sccm3H!! Where: x/sec=parameter representing the number of 2H+3H reactions required per second to generate 2 GWatt ex=10 to the x power MeV=Mega electron volt (energy) 2H=Deuterium 3H=Tritium GWatt=gigawatt (power, energy/time) 1 electronvolt=1.6021773×10−19 joule 1 watt=1 joule/second 1 u=1 unified atomic mass unit=1.660538921×10−24 gram 1 Mole of gas=6.022×1023 atoms or molecules=22.4 liters at standard conditions (23 C, 760 Torr (14.7 psi)) sccm=standard cubic centimeter per second flow rate (standard industry measure of flow rate) One of skill will appreciate that the location of the injectors for the two reactants is determined by their transport time to target. This interval is determined by their mass (resisting acceleration) and their ionization (producing force causing acceleration. The governing equation of rectilinear motion is:s=1/2at2 Where: s=distance from reactor induction to target a=acceleration due to the unbalanced force of the ionized reactants in the electromagnetic field t=the time of transport from the site of induction to the target The distance for the 2 reactants to the target can be the same, for example, due to the complimentary inverse relationship of mass to ionization—2/3 ratio for mass and 3/2 ratio for force due to relative ionization. One of skill will appreciate that the size of the target cathode should be related to the reactant flow for the reactor, the number of reaction nodes, the output of the reactor, and the size of the reactant nuclei. A configuration that offers a basis for establishing the physical reaction is a monolayer of reactant nuclei covering the faces of the electrode being impacted, recognizing that reactions adjacent to the target are anticipated either from same-side same-direction reactant collision or opposite-side opposite-direction reactant collision. This high density nuclear condition is unique to the teachings provided herein, and it produces a high reaction efficiency, as well as overcomes the inherent limitations in other unsatisfactory development paths at other facilities/programs. The true sizes of the atomic and nuclear species under discussion are dependent upon Bose Einstein Condensate behavior of the ionized bosons dependent on local momentum distribution (“temperature”), energy density conditions, and energy state. Referring to the calculated fuel flow rate:(x/sec)[2H(single molecule)+3H(single molecule)]=7.0926×1020/sec And, using the area of the reactants, and assuming a one reaction per second reactor frequency, gives:7.09261020[(2H(single molecule)+3H(single molecule)]×(2)×(9.6211×10−30 m2)=1.3648×10−8 m2 So for a 2 sided disc shaped target, the radius is ideally:r=√[(1.3648×10−8/2π)]=4.6606×10−5 m, or for a square targetI=·(1.3648×10−8)=1.1682×10−4 mWhere:Average estimated area of reactants is π[2×1.75 fm(1.75×10−15)]squared/4=π[3.5×10−15]squared/4=9.6211×10−30 m2. The diameter of the nucleus is in the range of 1.75 fm (1.75×10−15 m) for hydrogen (the diameter of a single proton) to about 15 fm for the heaviest atoms, such as uranium. These dimensions are much smaller than the diameter of the atom itself (nucleus+electron cloud), by a factor of about 23,000 (uranium) to about 145,000 (hydrogen). Moreover, it should be appreciated that, under the conditions in the reactor chamber during fusion, the target will/may evaporate after each reaction and replacement will/may be needed. The reaction chamber should be sized appropriately for the desired energy transfer, meaning that the exothermicity of the higher reaction efficiencies should be translated into the heat transfer surface area needed to maintain an efficient steady state process. Heat transfer through the high vacuum reaction chamber to the steam vessel is required to drive conventional turbines to generate electrical power. As such, it is the heat transfer through the chamber, maintenance and fabrication, and the reaction process that govern optimum chamber dimensions. Using conventional light water reactor general parameters as a reference, for example, a 2 GW Nuclear Fusion Power Plant Reactor would have 4 chamber modules each having a nominal 54 m2 surface area. Heat transfer levels into the chamber wall from the reaction will not be excessive resulting in chamber failure. A high vacuum reaction chamber failure mode could manifest as a leak—where the heat transfer medium, steam, or atmospheric air enters the chamber and prevents the high vacuum conditions required for the process.
050911422	abstract	Each of the zones of the ferrule (25) comprising a radially projecting deformed part (29) is deformed by inward bending, so as to extract the deformed part (29) from a corresponding cavity (22) of the end block. The locking sleeve (20) is then extracted by exerting a pulling force on this sleeve (20) in the axial direction (ZZ') of the guide tube (4). Preferably, the ferrule (25) consists of cylindrical segments separated by slits arranged along the generatrices of the ferrule (25). A tool (31) is used to fold the segments of the ferrule (25) and to grip these segments so as to extract the locking sleeve (20). Extraction is performed underwater in a fuel storage well.
	summary
047643323	claims	1. A method for the hydrotesting of an open-end pipe having an end wall, situated within a nuclear reactor pressure vessel, comprising: providing a pipe end sealing element having; a hollow tubular member, adapted to encircle the open end of the pipe, having an inwardly directed flange at one end thereof, said flange having an inner bevelled surface, and a threaded portion of the inner face of said hollow tubular member at the other end thereof adjacent the end of the pipe; a hollow conical bushing having an inner surface engageable with the outer surface of the pipe spaced from the open end thereof, said inner surface having an interior diameter and said bushing having means for varying said interior diameter, and said bushing having an outer bevelled surface complementary to the bevel of the bevelled surface of the flange of the hollow tubular member; a sealing plug insertable into the open end of said pipe with a portion thereof resting on the end wall of said pipe; and a threaded bolt threadably engageable with the threaded portion of the hollow tubular member; placing the hollow tubular member over the open end of the pipe, with the threaded portion thereof adjacent and beyond the open end thereof; inserting said hollow conical bushing into said hollow tubular member such that the same surrounds the outer wall of the pipe, with the bevelled surface of the conical bushing in sliding relationship to the bevelled surface on the flange of the hollow tubular member; placing said sealing plug on the open end of the pipe; engaging said threaded bolt with the threads of the hollow tubular member; advancing said threaded bolt towards said sealing plug to contact the same, while preventing rotation of the hollow tubular member, such that the hollow conical bushing is decreased in its interior diameter to grip the outer wall of the pipe, by axial sliding movement of the bevelled surface of the hollow tubular member, and finally hold the sealing plug in sealing contact with the end wall of the pipe; and injecting a fluid into said pipe from a location spaced from the end having the sealing element thereof to a predetermined pressure to determine the fluid resistant capability of the wall of said pipe. a hollow tubular member, adapted to encircle the open end of the pipe, having an inwardly directed flange at one end thereof, said flange having an inner bevelled surface, and a threaded portion on the inner face of said hollow tubular member at the other end thereof adjacent the end of the pipe; a hollow conical bushing having an inner surface engageable with the outer surface of the pipe spaced from the open end thereof, said inner surface having an interior diameter, and said bushing having means for varying said interior diameter, and said bushing having an outer bevelled surface complementary to the bevel of the bevelled surface of the flange of the hollow tubular member; a sealing plug insertable into the open end of said pipe, with a portion thereof resting on the end wall of said pipe; and a threaded bolt threadably engageable with the threaded portion of the hollow tubular member, whereby upon threading of the bolt into said hollow tubular member, the bevelled surface of the hollow tubular member slides along the bevelled surface of the conical bushing to decrease the interior diameter of the conical bushing and force the conical bushing into further engagement with the outer surface of the pipe, while forcing the sealing plug into sealing relationship with the end of the pipe. a hollow tubular member, adapted to encircle the open end of the pipe, having an inwardly directed flange at one end thereof, said flange having an inner bevelled surface, and a threaded portion on the inner face of said hollow tubular member at the other end thereof adjacent the end of the pipe; a hollow conical bushing, having a base end and an apical end, and a slot therein extending from said base end to said apical end, said bushing of variable interior diameter having an inner surface engageable with the outer surface of the pipe spaced from the open end thereof, and an outer bevelled surface complementary to the bevel of the bevelled surface of the flange of the hollow tubular member; a sealing plug having a cylindrical portion insertable into the open end of the pipe and a flange portion adapted to rest on the end wall of the pipe, said cylindrical portion having a groove formed in the outer wall thereof, and an O-ring engaged within said groove; and a threaded bolt threadably engageable with the threaded portion of the hollow tubular member, whereby upon threading of the bolt into said hollow tubular member, the bevelled surface of the hollow tubular member slides along the bevelled surface of the conical bushing to decrease the interior diameter of the conical bushing and force the conical bushing to further engagement with the outer surface of the pipe, while forcing the sealing plug into sealing relationship with the end of the pipe. a hollow tubular member, adapted to encircle the open end of the pipe, having an inwardly directed flange at one end thereof, said flange having an inner bevelled surface, and a threaded portion on the inner face of said hollow tubular member at the other end thereof adjacent the end of the pipe; a hollow conical bushing having an inner surface engageable with the outer surface of the pipe spaced from the open end thereof, said inner surface having an interior diameter, and said bushing having a base end and an apical end, and a slot therein extending from said base end to said apical end so as to enable decreasing of the interior diameter thereof, and said bushing having an outer bevelled surface complementary to the bevel of the bevelled surface of the flange of the hollow tubular member; a sealing plug insertable into the open end of said pipe, with a portion thereof resting on the end wall of said pipe; and a threaded bolt threadably engageable with the threaded portion of the hollow tubular member, whereby upon threading of the bolt into said hollow tubular member, the bevelled surface of the hollow tubular member slides along the bevelled surface of the conical bushing to decrease the interior diameter of the conical bushing and force the conical bushing into further engagement with the outer surface of the pipe, while forcing the sealing plug into sealing relationship with the end of the pipe. 2. The method as defined in claim 1 wherein said hollow conical bushing has a base end and an apical end, and a slot therein extending from said base end to said apical end so as to enable decreasing of the interior diameter thereof. 3. The method as defined in claim 1 wherein said hollow tubular member has means thereon for preventing rotation of said hollow tubular member resulting from threading of said bolt into said hollow tubular member. 4. The method as defined in claim 3 wherein said means for preventing rotation of said hollow tubular member comprises an aperture in the wall thereof adapted for insertion therein of a stop member. 5. The method as defined in claim 4 wherein said stop member is a rod insertable into the aperture in said wall. 6. The method as defined in claim 1 wherein said sealing plug has a cylindrical portion insertable into the open end of the pipe and a flange portion adapted to rest on the end wall of a pipe. 7. The method as defined in claim 6 wherein said cylindrical portion has a groove formed in the outer wall thereof, and an O-ring is engaged within said groove. 8. The method as defined in claim 1 wherein said threaded bolt has a head portion adapted for use in turning the bolt and a raised portion on the end thereof opposite the head portion adapted to contact said sealing plug upon threaded engagement of said bolt with said hollow tubular member containing said sealing plug therein. 9. A pipe end sealing element, for use in sealing a pipe for hydrotesting, which pipe has an open end and an end wall thereon, comprising; 10. The pipe end sealing element as defined in claim 9 wherein said hollow tubular element has means thereon for preventing rotation of said hollow tubular element resulting from threading of said bolt into said hollow tubular member. 11. The pipe end sealing element as defined in claim 10 wherein said means for preventing rotation of said hollow tubular element comprises an aperture in the wall thereof adapted for insertion therein of a stop member. 12. The pipe end sealing element as defined in claim 11 wherein said stop member is a rod insertable into the aperture in said wall. 13. The pipe end sealing element as defined in claim 1 wherein said sealing plug has a cylindrical portion insertable into the open end of the pipe and a flange portion adapted to rest on the end wall of the pipe. 14. The pipe end sealing element as defined in claim 13 wherein said cylindrical portion has a groove formed in the outer wall thereof, and an O-ring is engaged within said groove. 15. The pipe end sealing element as defined in claim 1 wherein said threaded bolt has a head portion adapted for turning the bolt and a raised portion on the end thereof opposite the head portion adapted to contact said sealing plug upon threaded engagement of said bolt with said hollow tubular element containing said sealing plug therein. 16. A pipe end sealing element, for use in sealing a pipe for hydrotesting, which pipe has an open end and an end wall thereon, comprising: 17. A pipe end sealing element, for use in sealing a pipe for hydrotesting, which pipe has an open end and an end wall thereon, comprising:
	claims	1. A nuclear fission reactor structure comprising a plurality of layers, wherein each layer of the plurality of layers includes:an inner segment body including an inner opening extending axially from a first side of the inner segment body to a second side of the inner segment body;an intermediate segment body radially outward of the inner segment body;an outer segment body radially outward of the intermediate segment body;a first interior interface separating the inner segment body and the intermediate segment body; anda second interior interface separating the intermediate segment body and the outer segment body,wherein, in a cross-sectional plan view in a plane perpendicular to the axially extending inner opening:the inner segment body includes a plurality of inner cladding arms having a first involute curve shape that spirally radiates outward from a first radially inner end adjacent to the inner opening to a first radially outer end at the first interior interface,the intermediate segment body includes a plurality of intermediate cladding arms having a second involute curve shape that spirally radiates outward from a second radially inner end adjacent to the first interior interface to a second radially outer end at the second interior interface, andthe outer segment body includes a plurality of outer cladding arms having a third involute curve shape that spirally radiates outward from a third radially inner end adjacent to the second interior interface to a third radially outer end at a radially outer surface of the outer segment body,wherein the inner cladding arms have opposing side surfaces extending from the first side of the inner segment body to the second side of the inner segment body, and wherein at least one protrusion projects outwardly from at least one opposing side surface, andwherein the protrusion has a top surface distal from the at least one opposing side surface for which the protrusion projects, and wherein, when assembled in the inner segment body with a first inner cladding arm immediately adjacent a second inner cladding arm, the top surface of a protrusion on the first inner cladding arm contacts an opposing side surface on the second inner cladding arm and forms a channel between the first inner cladding arm and the second inner cladding arm. 2. The nuclear fission reactor structure according to claim 1, wherein each of the first involute curve shape, the second involute curve shape, and the third involute curve shape have a different curvature. 3. The nuclear fission reactor structure according to claim 1, wherein each of the first involute curve shape, the second involute curve shape, and the third involute curve shape correspond to different portions of a continuous involute curve shape extending from the inner opening to the radially outer surface of the outer segment body. 4. The nuclear fission reactor structure according to claim 1, wherein, collectively, the first involute curve shape, the second involute curve shape, and the third involute curve shape form a continuous involute curve shape extending from the inner opening to the radially outer surface of the outer segment body. 5. The nuclear fission reactor structure according to claim 4, wherein a projection of a surface of the continuous involute curve shape extends across the first interior interface and the second interior interface and is coincident with each of a surface of one of the plurality of inner cladding arms, a surface of one of the plurality of intermediate cladding arms, and a surface of one of the plurality of outer cladding arms. 6. The nuclear fission reactor structure according to claim 1, wherein each of the first involute curve shape, the second involute curve shape, and the third involute curve shape correspond to different portions of a continuous involute curve shape extending from the inner opening to the radially outer surface of the outer segment body. 7. The nuclear fission reactor structure according to claim 6, wherein a projection of a surface of the continuous involute curve shape extends across the first interior interface and the second interior interface and is coincident with each of a surface of one of the plurality of inner cladding arms, a surface of one of the plurality of intermediate cladding arms, and a surface of one of the plurality of outer cladding arms. 8. The nuclear fission reactor structure according to claim 1, wherein each of the plurality of inner cladding arms, the plurality of intermediate cladding arms, and the plurality of outer cladding arms include a plurality of chambers. 9. The nuclear fission reactor structure according to claim 8, wherein the chambers in each cladding arm are separated from each other by a web. 10. The nuclear fission reactor structure according to claim 8, wherein each inner cladding arm has more chambers than each outer cladding arm. 11. The nuclear fission reactor structure according to claim 8, wherein the plurality of chambers is ten or less. 12. The nuclear fission reactor structure according to claim 8, wherein each intermediate cladding arm has the same number of chambers as each outer cladding arm. 13. The nuclear fission reactor structure according to claim 8, wherein the chambers include one of a fissionable fuel composition and a moderator material. 14. The nuclear fission reactor structure according to claim 13, wherein chambers at different locations along the cladding arm contain different fissionable fuel compositions. 15. The nuclear fission reactor structure according to claim 14, wherein chambers at different locations along the cladding arm contain different moderator material. 16. The nuclear fission reactor structure according to claim 13, wherein, when a fissionable fuel composition is located in a chamber, there is a space between at least a portion of one interior surface wall of the chamber and at least a portion of one exterior surface of a body formed of the fissionable fuel composition. 17. The nuclear fission reactor structure according to claim 1, wherein each protrusion extends along the at least one opposing side surface continuously from a first end oriented toward the first side of the inner segment body to a second end oriented toward the second side of the inner segment body. 18. The nuclear fission reactor structure according to claim 1, wherein each protrusion extends along the at least one opposing side surface discontinuously from a first end oriented toward the first side of the inner segment body to a second end oriented toward the second side of the inner segment body. 19. The nuclear fission reactor structure according to claim 1, wherein the inner segment body, the intermediate segment body, the outer segment body, the first interior interface, and second interior interface define a layer. 20. A nuclear fission reactor, comprising:a plurality of layers according to claim 19, wherein the plurality of layers are assembled into a nuclear fission reactor structure with a first end surface, a second end surface, and an outer side surface connecting the first end surface to the second end surface;a radial reflector positioned about the outer side surface of the nuclear fission reactor structure;a pressure vessel; anda coolant system in fluid communication with the nuclear fission reactor structure through openings in the pressure vessel. 21. The nuclear fission reactor according to claim 20, wherein the nuclear fission reactor structure has a cylindrical structure. 22. The nuclear fission reactor according to claim 20, wherein the coolant system is fluid-based or gas-based. 23. The nuclear fission reactor according to claim 20, wherein the first interior interface and the second interior interface in each of the plurality of layers include a plurality of secondary coolant channels that traverse the nuclear fission reactor structure from the first end to the second end. 24. A method of fabricating the nuclear fission reactor structure according to claim 1, the method comprising:manufacturing the inner segment body, segments of the intermediate segment body, and segments of the outer segment body, wherein each of the plurality of inner cladding arms, the plurality of intermediate cladding arms, and the plurality of outer cladding arms include a plurality of chambers;assembling the inner segment body, the segments of the intermediate segment body, and the segments of the outer segment body into a layer, wherein the segment bodies are assembled by one of welding and bonding;positioning one of a fissionable fuel composition and a moderator material in the plurality of chambers to form a fuel-loaded layer; andassembling a plurality of fuel-loaded layers into the nuclear fission reactor structure. 25. The method according to claim 24, wherein the inner segment body, segments of the intermediate segment body, and segments of the outer segment body are manufactured using an additive manufacturing process. 26. The method according to claim 24, including positioning the nuclear fission reactor structure within a radial reflector, wherein the nuclear fission reactor structure has a cylindrical shape. 27. A method of fabricating the nuclear fission reactor structure according to claim 1, the method comprising:manufacturing a layer including the inner segment body, the intermediate segment body, and the outer segment body as a unitary structure, wherein each of the plurality of inner cladding arms, the plurality of intermediate cladding arms, and the plurality of outer cladding arms include a plurality of chambers;positioning one of a fissionable fuel composition and a moderator material in the plurality of chambers to form a fuel-loaded layer; andassembling a plurality of fuel-loaded layers into the nuclear fission reactor structure. 28. The method according to claim 27, wherein the unitary structure of the inner segment body, the intermediate segment body, and the outer segment body are manufactured using an additive manufacturing process. 29. The method according to claim 27, including positioning the nuclear fission reactor structure within a radial reflector, wherein the nuclear fission reactor structure has a cylindrical shape.
051851040	claims	1. A method of treatment of a high-level radioactive waste of a nitric acid solution obtained as an extraction residue in reprocessing of spent nuclear fuels, said method consisting essentially of: calcining the nitric acid solution to evaporate moisture and nitric acid in the solution to thereby obtain a calcined radioactive waste; heating the calcined radioactive waste to a temperature ranging from about 500.degree. to about 3,000.degree. C. to vaporize a portion of the elements contained in the radioactive waste, the respective elements being vaporized in the form of metal or oxide thereof; and obtaining as a residue, a volume-reduced high-level radioactive solid. calcining the nitric acid solution to evaporate moisture and nitric acid in the solution to thereby obtain a calcined radioactive waste; heating the calcined radioactive waste in the presence of a reducing agent at a temperature ranging from about 500.degree. to about 3,000.degree. C. to vaporize a portion of the elements contained in the radioactive waste, the respective elements being vaporized in the form of metal; and obtaining as a residue, a volume-reduced high-level radioactive solid. calcining the nitric acid solution to evaporate moisture and nitric acid in the solution to thereby obtain a calcined radioactive waste; heating the calcined radioactive waste to a temperature ranging from about 500.degree. to about 3,000.degree. C. to vaporize a first portion of the elements contained in the radioactive waste, the respective elements being vaporized in the form of metal or oxide thereof; heating the remaining radioactive waste in the presence of a reducing agent at a temperature ranging from about 500.degree. to about 3,000.degree. C. to vaporize a second portion of the elements contained in the remaining radioactive waste, the respective elements being vaporized in the form of metal; and obtaining as a residue, a volume-reduced high-level radioactive solid. calcining the nitric acid solution to evaporate moisture and nitric acid in the solution to thereby obtain a calcined radioactive waste; heating the calcined radioactive waste in the presence of a reducing agent at a temperature ranging from about 500.degree. to about 3,000.degree. C. to vaporize a portion of the elements contained in the radioactive waste, the respective elements being vaporized in the form of metal and including cesium, strontium or cesium and strontium; and obtaining as a residue, a volume-reduced high-level radioactive solid containing therein no cesium and strontium. 2. The method according to claim 1, which further comprises cooling stepwise the resultant vapor at different temperatures each corresponding to the sublimation or boiling point of each metal or each oxide thereof to separately collect the respective elements. 3. A method of treatment of a high-level radioactive waste of a nitric acid solution obtained as an extraction residue in reprocessing of spent nuclear fuels, said method consisting essentially of: 4. The method according to claim 3, which further comprises cooling stepwise the resultant vapor at different temperatures each corresponding to the sublimation or boiling point of each metal to separately collect the respective elements. 5. A method of treatment of a high-level radioactive waste of a nitric acid solution obtained as an extraction residue in reprocessing of spent nuclear fuels, said method consisting essentially of: 6. The method according to claim 5, which further comprises cooling stepwise the resultant vapor of the first portion of the elements at different temperatures each corresponding to the sublimation or oiling point of each metal or each oxide thereof to separately collect the first portion of the elements, and cooling stepwise the resultant vapor of the second portion of the elements at different temperatures each corresponding to the sublimation or boiling point of each metal to separately collect the second portion of the elements. 7. A method of treatment of a high-level radioactive waste of a nitric acid solution obtained as an extraction residue in reprocessing of spent nuclear fuels and containing cesium, strontium or cesium and strontium, said method consisting essentially of:
052395669	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a mirror composed of multi-layered film for use in a soft X-ray region, and more particularly a multi-layered mirror adapted for use in a soft X-ray microscope for organism observation. 2. Related Background Art The refractive index of a substance in the X-ray region is represented by n=1-.delta.-ik wherein .delta. and k are real numbers much smaller than unity (the imaginary part ik represents the X-ray absorption). For this reason, lenses based on refractive action, as in the visible wavelength region, cannot be utilized in the X-ray region. Consequently there is utilized a reflective optical system. However, since the reflectivity is very low at an incident angle smaller than the totally reflecting limit angle .theta..sub.c (about 5.degree. or smaller for a wavelength of 25 .ANG.), there is employed a multi-layered mirror having a plurality of (for example several hundred) reflecting planes formed by laminating layers of substances of a combination showing a high amplitude reflectivity at the interface and regulating the thicknesses of said layers in such a manner that the reflected waves mutually match in phase according to the optical interference theory. More specifically, such multi-layered mirror can be obtained by alternately laminating a substance showing a larger difference in refractive index from vacuum (refractive index=1) and another substance showing a smaller difference. There are conventionally known certain examples of combinations of such substances, such as W (tungsten)/C (carbon) and Mo (molybdenum)/Si (silicon), and such mirrors have been prepared by thin film forming technologies such as sputtering, vacuum evaporation, CVD etc. The wavelength of X-ray employed for observation of living organism in a soft X-ray microscope is selected in a region called "water window" wherein proteins and water show a large difference in absorption coefficients as shown in FIG. 6, namely a region between the K absorption edge (23 .ANG.) of oxygen and the K absorption edge (44 .ANG.) of carbon. A wavelength of 25 .ANG., close to the absorption edge of oxygen showing a smaller absorption coefficient, is preferred in order to enable observation of a thicker specimen. Also the periodic thickness d (combined thickness of a pair of said substance having a larger difference in refractive index from vacuum and said substance having a smaller difference), shown in FIG. 1, has to approximately satisfy the Bragg's law 2d.multidot.sin .theta.=.lambda., wherein .lambda. and .theta. are respectively the wavelength of X-ray and the angle between the mirror surface and the incident X-ray. The conventional multi-layered mirror has been designed to reflect the X-ray in said "water window" region by suitably selecting said periodic thickness and said angle of the incident X-ray to the mirror surface. In a soft X-ray microscope or the like, the size of the multi-layered mirror can be made smaller and the freedom in the optical system design becomes larger, if the X-ray is made to enter the mirror as perpendicularly as possible. With such mode of entry, the microscope itself can be made more compact. Stated differently, said angle of the incident X-ray to the mirror surface should preferably be as large as possible. However, as will be apparent from the foregoing Bragg's law, the periodic thickness has to be made smaller in order to introduce the X-ray, within the above-mentioned wavelength region, into the multi-layered mirror in a state as close to the perpendicular entry as possible. For example, the periodic thickness has to be 24 .ANG. in order to introduce the X-ray of a wavelength of 33.7 .ANG. with an angle of 45.degree. to the multi-layered mirror. Thus, there has been a need for a multi-layered mirror having a high reflectivity and a reduced periodic thickness. Nickel has a large difference from vacuum in refractive index in the above-mentioned wavelength region, and is therefore expected to provide a high reflectivity. In practice, however, in a multi-layered film employing nickel, it has been experimentally confirmed that the reflectivity rapidly drops when the thickness of nickel layers (or periodic thickness of the mirror) is reduced. FIG. 3 shows the relationship between the periodic thickness and the ratio of measured reflectivity to calculated reflectivity when X-ray of a wavelength of 1.54 .ANG. is reflected by a multi-layered film consisting of nickel/silicon oxide. Said multi-layered film provides a reflectivity corresponding to 80% of the calculated value at a periodical thickness of 60 .ANG., but rapidly loses the reflectivity with the reduction in the periodical thickness, and becomes totally incapable of reflection when the periodical thickness reaches 30 .ANG.. Thus even the multi-layered film employing nickel loses reflectivity with the reduction in the periodical thickness, and is incapable of providing a high reflectivity in a state of a large angle of the incident X-ray to the mirror surface, in the aforementioned "water window" wavelength region. SUMMARY OF THE INVENTION The object of the present invention is to solve the above-mentioned problems. The above-mentioned object can be attained, according to the present invention, by a multi-layered mirror consisting of alternately laminated layers of a substance showing a larger difference in refractive index in the soft X-ray region from vacuum (refractive index=1) and another substance showing a smaller difference, wherein said substance of the larger difference in refractive index is composed of a nickel-chromium alloy containing chromium in 5 wt. % or larger, or pure chromium. FIGS. 4A and 4B show the optical constants .delta., k of nickel and chromium as a function of the wavelength of soft X-ray (20-50 .ANG.). As will be apparent from these charts, nickel shows a larger difference in refractive index from vacuum, because of larger values of .delta. and k. Nickel is therefore anticipated to provide a higher reflectivity in this wavelength region. However, as explained above, the multi-layered film employing nickel shows a rapid decrease in reflectivity, when the periodic thickness of the multi-layered film is reduced. FIG. 3 shows the relationship between the periodic thickness and the ratio of measured reflectivity to calculated reflectivity, for X-ray of a wavelength of 1.54 .ANG., in multi-layered films consisting of nickel/silicon oxide, nickel-chromium alloy (nickel 80 wt. %: chromium 20 wt. %)/silicon oxide, and chromium/silicon oxide. It will be understood that the deterioration in the reflectivity ratio, resulting with the reduction in the periodic thickness, can be significantly alleviated when nickel is replaced by nickel-chromium alloy or chromium. For soft X-ray within the wavelength range of 20-50 .ANG., the optical constants of nickel-chromium alloy assume values between those of pure nickel and those of pure chromium, depending on the mixing ratio of nickel and chromium. More specifically, said optical constants become closer to those of pure nickel when the proportion of chromium is low, and those of pure chromium when said proportion is high. Consequently, in comparison with a multi-layered film employing pure chromium, that employing a nickel-chromium alloy can provide a higher reflectivity with a fewer number of layers. FIG. 5 shows the reflectivity to soft X-ray of a wavelength of 33.7 .ANG., as a function of chromium content (in weight percent) in said alloy. The multi-layered film employed in this measurement was prepared by alternate laminations of 50 layers each of a nickel-chromium alloy with a thickness of 20 .ANG., and vanadium oxide with a thickness of 20 .ANG. as a substance with a smaller difference in refractive index from vacuum. As will be apparent from FIG. 5, addition of chromium into nickel provides a higher reflectivity, in comparison with the case of pure nickel. It is therefore possible to prevent the abrupt decrease in the ratio of reflectivity resulting from the reduction in the periodic thickness, by employing a nickel-chromium alloy or pure chromium, as the substance showing a large difference in refractive index from vacuum. Also as will be understood from FIG. 5, the reflectivity can be made higher than in the case of pure chromium through the use of an alloy in which the content of chromium in nickel is 5 % or higher. As explained in the foregoing, the present invention provides a multi-layered mirror which provides a high reflectivity even when the periodic thickness is reduced, by employing a nickel-chromium alloy or pure chromium as the substance showing a large difference in refractive index from vacuum. It is therefore possible to construct an optical system in which the soft X-ray can be are perpendicularly introduced into said multi-layered mirror, and to thus increase the freedom of designing of the optical system. It is also possible to reduce the dimension of the multi-layered mirror, thereby facilitating size and weight reduction of the equipment employing such multi-layered mirror. The present invention is applicable not only in the soft X-ray microscope for observation of living organism but also to any optical equipment employed in the soft X-ray region, such as X-ray microscopes for other purposes, X-ray lithographic equipment, X-ray telescope, X-ray laser or the like.
	summary
	description	This application generally relates to control of nuclear reactions and more particularly relates to a traveling wave nuclear fission reactor, fuel assembly, and a method of controlling burnup therein. It is known, in an operating nuclear fission reactor, that neutrons of a known energy are absorbed by nuclides having a high atomic mass. The resulting compound nucleus separates into fission products that include two lower atomic mass fission fragments and also decay products. Nuclides known to undergo such fission by neutrons of all energies include uranium-233, uranium-235 and plutonium-239, which are fissile nuclides. For example, thermal neutrons having a kinetic energy of 0.0253 eV (electron volts) can be used to fission U-235 nuclei. Thorium-232 and uranium-238, which are fertile nuclides, will not undergo induced fission, except with fast neutrons that have a kinetic energy of at least 1 MeV (million electron volts). The total kinetic energy released from each fission event is about 200 MeV. This kinetic energy is eventually transformed into heat. Moreover, the fission process, which starts with an initial source of neutrons, liberates additional neutrons as well as transforms kinetic energy into heat. This results in a self-sustaining fission chain reaction that is accompanied by continued energy release. A traveling wave Pyrotron for continuous operation is disclosed in U.S. Pat. No. 3,093,569, issued Jun. 11, 1963 in the names of Richard F. Post, et al. and titled “Traveling Wave Pyrotron.” This patent discloses a continuous operating reactor or device for increasing the energy and density of plasma and conducting nuclear reactions therein. An object of the invention is to provide a Pyrotron where traveling magnetic waves are employed to accomplish trapping, heating and energy recovery of charged particles within individual containment zones, each of which progresses along the machine with time. However, this patent does not appear to disclose a traveling wave nuclear fission reactor, fuel assembly, and a method of controlling burnup therein, as described and claimed herein. U.S. Pat. No. 3,799,839 issued Mar. 6, 1974 in the names of David L. Fischer, et al. and titled “Reactivity And Power Distribution Control Of Nuclear Reactor” discloses a spatial distribution, amount, density and configuration of burnable poison to control a predetermined amount of excess reactivity and to maintain a constant or stationary power distribution during the operating cycle of a nuclear reactor core. According to this patent, it is an object of the invention to provide an arrangement of burnable poison in a nuclear reactor core which will provide a substantially stationary power distribution in the core throughout the period of the operating cycle. Also, according to this patent, other objects are achieved in accordance with the invention by determining consistent power and concomitant reactivity distributions for the operating cycles: by determining the resulting excess local reactivity, and by providing burnable poison in amount, density and configuration, spatially distributed to substantially match the changes in excess local reactivity throughout the period of the operating cycle. However, this patent does not appear to disclose a traveling wave nuclear fission reactor, fuel assembly, and a method of controlling burnup therein, as described and claimed herein. U.S. Pat. No. 3,489,646 issued Jan. 13, 1970 in the names of Jean Paul Van Dievoet, et al. and titled “Method of Pulsating or Modulating a Nuclear Reactor” relates to a method of pulsating or modulating the operation of a nuclear reactor. This patent discloses modulating the reactor by periodically varying the neutron flux density. According to this patent, operation of a nuclear reactor is controlled by moving one or more structures containing at least at certain localities, an amount of neutron-active substance, at a place outside the nuclear fission region of the reactor, and thereby modify, in dependence upon the speed of the structure, a neutron flow issuing from the reactor core. The specimens of neutron-active material which thus modify the reactivity of the reactor system from the outside may be neutron-generating and/or neutron-influencing material, such as fissionable material, reflector material or other neutron-influencing substance. However, this patent does not appear to disclose a traveling wave nuclear fission reactor, fuel assembly, and a method of controlling burnup therein, as described and claimed herein. None of the art recited hereinabove appears to disclose a traveling wave nuclear fission reactor, fuel assembly, and a method of controlling burnup therein, as described and claimed herein. Therefore, what are needed are a traveling wave nuclear fission reactor, fuel assembly, and a method of controlling burnup therein, as described and claimed herein. According to an aspect of this disclosure, there is provided a method of controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux, comprising modulating the neutron flux emitted by the traveling wave nuclear fission reactor. According to another aspect of this disclosure, there is provided a method of controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux, comprising modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. According to a further aspect of this disclosure, there is provided a traveling wave nuclear fission reactor, comprising a nuclear reactor core; and a nuclear fission reactor fuel assembly disposed in the reactor core, the nuclear fission reactor fuel assembly being configured to achieve a burnup value at or below a predetermined burnup value. According to an additional aspect of this disclosure, there is provided a traveling wave nuclear fission reactor, comprising a nuclear reactor core capable of producing a burnfront therein; a nuclear fission reactor fuel assembly disposed in the nuclear reactor core; a neutron interactive material disposed in the nuclear fission reactor fuel assembly; and a control system configured to control disposition of the nuclear interactive material in response to a parameter associated with the burnfront. According to yet another aspect of this disclosure, there is provided a traveling wave nuclear fission reactor capable of controlling burnup therein, comprising a reactor pressure vessel; a nuclear fission reactor fuel assembly sealingly disposed in the pressure vessel, the nuclear fission reactor fuel assembly including a neutron interactive material arranged in a predetermined loading pattern; and a removable nuclear fission igniter capable of being disposed in neutronic communication with the neutron interactive material, the nuclear fission igniter capable of igniting a deflagration wave burnfront traveling through the neutron interactive material. A feature of the present disclosure is the provision of neutron absorber material in the form of a control rod, reflector, or neutron emitting material or other absorber material that enhances absorption at a location relative to a deflagration wave burnfront. In addition to the foregoing, various other method and/or device aspects are set forth and described in the teachings such as text (e.g., claims and/or detailed description) and/or drawings of the present disclosure. The foregoing is a summary and thus may contain simplifications, generalizations, inclusions, and/or omissions of detail; consequently, those skilled in the art will appreciate that the summary is illustrative only and is not intended to be in any way limiting. In addition to the illustrative aspects, embodiments, and features described above, further aspects, embodiments, and features will become apparent by reference to the drawings and the following detailed description. In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, similar symbols typically identify similar components, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented herein. In addition, the present application uses formal outline headings for clarity of presentation. However, it is to be understood that the outline headings are for presentation purposes, and that different types of subject matter may be discussed throughout the application (e.g., device(s)/structure(s) may be described under process(es)/operations heading(s) and/or process(es)/operations may be discussed under structure(s)/process(es) headings; and/or descriptions of single topics may span two or more topic headings). Hence, the use of the formal outline headings is not intended to be in any way limiting. Moreover, the herein described subject matter sometimes illustrates different components contained within, or connected with, different other components. It is to be understood that such depicted architectures are merely exemplary, and that in fact many other architectures may be implemented which achieve the same functionality. In a conceptual sense, any arrangement of components to achieve the same functionality is effectively “associated” such that the desired functionality is achieved. Hence, any two components herein combined to achieve a particular functionality can be seen as “associated with” each other such that the desired functionality is achieved, irrespective of architectures or intermedial components. Likewise, any two components so associated can also be viewed as being “operably connected”, or “operably coupled,” to each other to achieve the desired functionality, and any two components capable of being so associated can also be viewed as being “operably couplable,” to each other to achieve the desired functionality. Specific examples of operably couplable include but are not limited to physically mateable and/or physically interacting components, and/or wirelessly interactable, and/or wirelessly interacting components, and/or logically interacting, and/or logically interactable components. In some instances, one or more components may be referred to herein as “configured to,” “configurable to,” “operable/operative to,” “adapted/adaptable,” “able to,” “conformable/conformed to,” etc. Those skilled in the art will recognize that “configured to” can generally encompass active-state components and/or inactive-state components and/or standby-state components, unless context requires otherwise. Some considerations regarding various embodiments disclosed herein are given by way of overview but are not to be interpreted as limitations. Also, some of the embodiments disclosed herein reflect attainment of all of the considerations discussed below. On the other hand, some other embodiments disclosed herein reflect attainment of selected considerations, and need not accommodate all of the considerations discussed hereinbelow. Portions of the following discussion include information excerpted from a paper titled “Completely Automated Nuclear Power Reactors for Long-Term Operation: III. Enabling Technology 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). As previously mentioned, for every neutron that is absorbed in a fissile nuclide which leads to a fission event, more than one neutron is liberated until the fissile nuclei are depleted. This phenomenon is used in a commercial nuclear reactor to produce continuous heat that, in turn, is beneficially used to generate electricity. However, a consideration in reactor design and operation is heat damage to reactor structural materials due to “peak” temperature (i.e., hot channel peaking factor) which occurs due to a combination of uneven neutron flux, coolant flow, fuel composition and power distribution in the reactor. Heat damage results if the peak temperature exceeds material limits. This can happen regardless of the extent of burn-up (i.e., cumulative amount of energy generated per unit mass of fuel), which is usually expressed in units of megawatt-days per metric tonne of heavy metal fuel (MWd/MTHM) or gigawatt-days per metric tonne of heavy metal fuel (GWd/MTHM). A “reactivity change” (i.e., change in the responsiveness of the reactor) may be produced because of fuel burnup. More specifically, “reactivity change” is related to the relative ability of the reactor to produce more or less neutrons than the exact amount to sustain the critical chain reaction. Responsiveness of a reactor is typically characterized as the time derivative of a reactivity change causing the reactor to increase or decrease in power exponentially where the time constant is known as the reactor period. In this regard, control rods made of neutron absorbing material are typically used to adjust and control the changing reactivity and reactor responsiveness. Such control rods are reciprocated in and out of the reactor core to variably control neutron absorption and thus the neutron flux level and reactivity in the core. The neutron flux level is depressed in the vicinity of the control rod and potentially higher in areas remote from the control rod. Thus, the neutron flux is not uniform across the reactor core. This results in higher fuel burnup in those areas of higher flux. Also, it may be appreciated by a person of ordinary skill in the art of nuclear power production that flux and power density variations are due to many factors. Proximity to a control rod may or may not be the primary factor. For example, the flux typically drops significantly at core boundaries with no nearby control rod. These effects, in turn, may cause overheating or high temperatures in those areas of higher flux. Such peak temperatures may undesirably reduce the operational life of structures subjected to such peak temperatures by altering the mechanical properties of the structures. Also, reactor power density, which is proportional to the product of the neutron flux and the fissile fuel concentration, is limited by the ability of core structural materials to withstand such high temperatures without damage. Therefore, it is desirable to avoid structural damage due to high temperatures caused by high fuel burnup. Another consideration in reactor design and operation is irradiation damage to structural materials contained in the nuclear reactor core due to high fuel burnup. Such irradiation damage may be expressed in terms of displacements per atom (DPA), which includes information on the response of the material (i.e., displaced atoms), as well as the fast neutron fluence to which the material was exposed. DPA is proportional to burnup and is a calculated, representative measure of radiation damage which accounts for not only the dose and type of irradiation, but also includes a measure of the material's response to the irradiation. In this regard, some structural materials used in reactor core structures may become embrittled when exposed to neutrons released during the fission process. It is desirable to maintain such irradiation damage to reactor structural materials within known limits in order to ensure structural integrity and safe operation of the reactor. Therefore, referring to FIG. 1, by way of example only and not by way of limitation, there is shown a nuclear fission reactor arrangement, generally referred to as 10, to address the problems recited hereinabove. Nuclear fission reactor arrangement 10 generates electricity that is to be transmitted over a plurality of transmission lines (not shown). Reactor arrangement 10 alternatively may be used to conduct tests to determine effects of neutron flux on reactor materials. Referring again to FIG. 1, reactor arrangement 10 comprises a nuclear fission reactor, generally referred to as 20, that includes a plurality of generic nuclear fission reactor fuel assemblies, generally referred to as 30 (only one of which is shown), disposed within a reactor pressure vessel 40, which in turn may be housed within a containment structure (not shown). By way of illustration only and not by way of limitation, exemplary embodiments of generic fuel assembly 30 are disclosed hereinbelow. Generic fuel assembly 30 may be surrounded by a neutron multiplier or reflector material (not shown) and a radiation shield (also not shown). In this case, the reflector material reduces neutron leakage from fuel assembly 30. An additional function of the reflector material is to substantially reduce the fast neutron fluence seen by the outer portions of fuel assembly 30, such as its radiation shield, structural supports and containment structure. It also influences the performance of generic fuel assembly 30, so as to improve the breeding efficiency and the specific power in the outermost portions of generic fuel assembly 30. The radiation shield, on the other hand, further protects the biosphere from unintended release of radiation from generic fuel assembly 30. Referring yet again to FIG. 1, a primary coolant loop 50 carries heat from Generic fuel assembly 30 to a steam generating heat exchanger 60. Primary loop 50 may be made from any suitable material, such as stainless steel. Thus, if desired, primary loop 50 may be made from ferrous alloys, non-ferrous alloys, zirconium-based alloys or other structural materials or composites. The coolant carried by primary loop 50 may be a noble gas or mixtures thereof. Alternatively, the coolant may be other fluids such as water (H2O), or gaseous or supercritical carbon dioxide (CO2). As another example, the coolant may be a liquid metal such as sodium (Na) or lead (Pb) or alloys, such as lead-bismuth (Pb—Bi). Further, the coolant may be an organic-based coolant, such as a polyphenyl or a fluorocarbon. As the coolant carried by primary loop 50 passes through steam generating heat exchanger 60, the coolant will give-up its heat to a working fluid (not shown) residing in heat exchanger 60. The working fluid will vaporize to steam, when the working fluid is water. In this case, the steam travels into a secondary loop 70 that is isolated from primary loop 50 and that is coupled to a turbine-generator set 80a and 80b. Hence, heat exchanger 60 transfers heat to the working fluid in heat exchanger 60 and secondary loop 70 to generate steam that is provided as the working fluid for rotating turbine-generator set 80a and 80b. Turbine-generator set 80a and 80b generates electricity as it rotates, in a manner well understood in the art of electricity production from steam. A condenser 90 may be suitably coupled to turbine-generator set 80a and 80b for condensing exhaust steam from turbine-generator set 80a and 80b from a vapor state to a liquid state. Referring again to FIG. 1, a pump 100 is coupled to secondary loop 70 and is in fluid communication with the working fluid carried by secondary loop 70 for pumping the liquefied working fluid from condenser 90 heat exchanger 60. Moreover, a pump 110 is coupled to primary loop 50 and is in fluid communication with the reactor coolant carried by primary loop 50 for pumping the reactor coolant through primary loop 50. Primary loop 50 carries the reactor coolant from generic fuel assembly 30 to heat exchanger 60. Also, primary loop 50 carries the coolant from heat exchanger 60 to pressure vessel 40. Pump 110 circulates the reactor coolant through primary loop 50, including generic fuel assembly 30 and heat exchanger 60 in order to remove heat generated by fuel assembly 30 during reactor operation or to remove residual decay heat when reactor 20 is not operating. Removing heat from generic fuel assembly 30 reduces the risk that generic fuel assembly 30 will overheat, which is highly undesirable. Referring now to FIGS. 2 and 3, generic fuel assembly 30 suitably utilizes a fast neutron spectrum because the high absorption cross-section of fission products for epithermal to thermal neutrons does not permit utilization of more than a small amount of thorium or of the more abundant uranium isotope, U238, in uranium-fueled embodiments, because of neutron absorption by fission products. As best seen in FIG. 2, 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 (approximately 0.1 eV) energies, but are comparatively negligible above the resonance capture region (between approximately 3 and 300 eV). Thus, operating with a fast neutron spectrum when attempting to realize a high-gain fertile-to-fissile breeder can help to preclude neutron losses to fission products that build-up within the core during operation. 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 a negligible extent. Those in the central portions of the burn-waves of embodiments of generic fuel assembly 30 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. 3, 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. 3. The neutron spectrum of embodiments of generic fuel assembly 30 peaks in the ≧105 eV neutron energy region. The lower portion of FIG. 3 contains the ratio of these cross-sections versus 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—Pu239beta decay-chain upon neutron capture by U238). Thus, it can be seen that losses to radiative capture on fission products are comparatively minimized for a reactor having a fast spectrum. Turning now to FIGS. 4 and 5, generic fuel assembly 30 comprises fissile and/or fertile material, which may take the form of a plurality of elongate nuclear fission reactor fuel rods 150 (only some of which are shown) arranged in a predetermined fuel loading pattern. Exemplary embodiments of generic fuel assembly 30 are disclosed hereinbelow. Fuel rods 150 are sealingly contained within a leak-tight enclosure 155. Each fuel rod 150 has nuclear fuel 160 disposed therein, which nuclear fuel 160 is sealingly surrounded by a fuel rod cladding material 170. An average burnup value for fuel assembly 30 is limited by cladding material 170, which is the most pertinent structural material within fuel assembly 30. Nuclear fuel 160 comprises the afore-mentioned fissile nuclide, such as uranium-235, uranium-233 or plutonium-239. Alternatively, nuclear fuel 160 may comprise a fertile nuclide, such as thorium-232 and/or uranium-238 which will be transmuted during the fission process into the fissile nuclides mentioned immediately hereinabove. A further alternative is that nuclear fuel 160 may comprise a predetermined mixture of fissile and fertile nuclides. By way of example only, and not by way of limitation, nuclear fuel 160 may be made from an oxide selected from the group consisting essentially of uranium monoxide (UO), uranium dioxide (UO2), thorium dioxide (ThO2) (also referred to as thorium oxide), uranium trioxide (UO3), uranium oxide-plutonium oxide (UO—PuO), triuranium octoxide (U3O8) and mixtures thereof. Alternatively, nuclear fuel 160 may substantially comprise uranium alloyed with other metals, such as, but not limited to, zirconium or thorium metal alloyed or unalloyed. As yet another alternative, nuclear fuel 160 may substantially comprise a carbide of uranium (UCx) or a carbide of thorium (ThCx). For example, nuclear fuel 160 may be made from a carbide selected from the group consisting essentially of uranium monocarbide (UC), uranium dicarbide (UC2), uranium sesquicarbide (U2C3), thorium dicarbide (ThC2), thorium carbide (ThC) and mixtures thereof. As another non-limiting example, nuclear fuel 160 may be made from a nitride selected from the group consisting essentially of uranium nitride (U3N2), uranium nitride-zirconium nitride (U3N2Zr3N4), uranium-plutonium nitride ((U—Pu)N), thorium nitride (ThN), uranium-zirconium alloy (UZr) and mixtures thereof. Fuel rod cladding material 170, which sealingly surrounds nuclear fuel 160, may be a suitable zirconium alloy, such as ZIRCOLOY™ (trademark of the Westinghouse Electric Corporation), which has known resistance to corrosion and cracking. Cladding material 170 may be other materials, as well, such as ferritic martensitic steels. Referring to FIGS. 4 and 6, generic fuel assembly 30 further comprises neutron absorber material, which may take the form of a plurality of elongate neutron absorber or control rods 180 (only some of which are shown) with an associated control rod cladding 190. Control rods 180 are capable of introducing negative reactivity into generic fuel assembly 30. Control rods 180 may be in the form of “part-length” control rods 192 (only some of which are shown) and/or “full-length” control rods 194 (only some of which are shown). Full-length control rods 194 are suitably positioned parallel to fuel rods 150 and extend the entire length of fuel rods 150 when fully inserted into an enclosure 155. Part-length control rods 192 are also suitably positioned parallel to fuel rods 150, but do not extend the entire length of fuel rods 150 when fully inserted into enclosure 155. There may be any number of such part-length and full-length control rods depending on the neutron flux shaping design requirements for fuel assembly 30. A purpose of full-length control rods 192 is to reduce the rate of or stop the fission process occurring within generic fuel assembly 30 such as before decommissioning of reactor arrangement 10. Moreover, control rod and/or fuel rod configurations may deviate from the classic rod assembly type configurations mentioned immediately hereinabove. For example, plate type fuel may be used. Additionally, the fuel rods may be perpendicular (or at any other angle) to the direction of burn. Still referring to FIGS. 4 and 6, each control rod 180 comprises a suitable neutron absorber material 200 having an acceptably high neutron capture cross-section. In this regard, absorber material 200 may be a metal or metalloid selected from the group consisting essentially of lithium, silver, indium, cadmium, boron, cobalt, hafnium, dysprosium, gadolinium, samarium, erbium, europium and mixtures thereof. Alternatively, absorber material 200 may be a compound or alloy selected from the group consisting essentially of silver-indium-cadmium alloy, boron carbide, zirconium diboride, titanium diboride, hafnium diboride, gadolinium titanate, dysprosium titanate and mixtures thereof. Additionally, fuel rods that have been burned and have a high fission product concentration may be used as part of the control. By way of example only, and not by way of limitation, each of such controls rods 180 is, for example, vertically slidably movable inside respective ones of a plurality of control rod guide tubes (not shown) that were previously fixed within generic fuel assembly 30, such as during fabrication of generic fuel assembly 30. A purpose of part-length control rods 194 is to fine-tune the neutron flux within generic fuel assembly 30, so as to achieve a more precise burn-up of the fuel within generic fuel assembly 30. Referring again to FIGS. 4 and 6, control rods 180 are selectively operable, such as by means of respective ones of a plurality of drive motors 210 controlled by a controller (not shown). Each drive motor 210 engages its respective control rod 180 when electrical power is supplied to motor 210 and suitably disengages control rod 180 when electrical power is not supplied to motor 210, such as during a loss of power incident. Thus, if a loss of power incident occurs, motor 210 will disengage control rod 180, so that control rod 180 will vertically slidably drop into generic fuel assembly 30 along an interior of the previously mentioned guide tube due to force of gravity. In this manner, control rods 180 will controllably supply negative reactivity to generic fuel assembly 30. Thus, generic fuel assembly 30, by means of control rods 180, provides a reactivity management capability in the event of a loss of power incident, without reactor operator control or intervention. Referring to FIG. 7, generic fuel assembly 30 may further comprise a neutron multiplier or reflector, which may take the form of a plurality of elongate neutron reflector rods 220 sealingly housed in a reflector rod cladding 230. Reflector rods 220 cause elastic scattering of neutrons and are thus intended to “reflect” neutrons. Due to such elastic scattering of neutrons, reflector rods 220 are capable of introducing positive reactivity into fuel assembly 30 by decreasing the neutron leakage from generic fuel assembly 30. In this regard, each reflector rod 220 comprises a suitable neutron reflector material 240 having a suitable probability for neutron scattering. In this regard, reflector material 240 may be a material selected from the group consisting essentially of beryllium (Be), lead alloys, tungsten (W), vanadium (V), depleted uranium (U), thorium (Th) and mixtures thereof. Reflector material 240 may also be selected from a wide variety of steel alloys. It should be appreciated that the fissile and fertile materials that are contemplated for use in generic fuel assembly 30 also have high elastic scattering cross sections. Returning to FIG. 4, generic fuel assembly 30 further comprises a comparatively small and removable nuclear fission igniter 245 that includes moderate isotopic enrichment of nuclear fissionable material, such as, without limitation, U233, U235 or Pu239, suitably centered in enclosure 155 along a vertical axis 247a. Igniter 245 may be disposed at an end of enclosure 155 rather than being centered in enclosure 155, if desired. Neutrons are released by igniter 245. The neutrons that are released by igniter 245 are captured by the fissile and/or fertile material within fuel rods 150 to initiate the previously mentioned fission chain reaction. Igniter 245 may be removed once the chain reaction becomes self-sustaining, if desired. It will be understood that the teachings herein describe a traveling wave nuclear fission reactor. The basic principles of such a traveling wave nuclear fission reactor is disclosed in more detail in co-pending U.S. patent application Ser. No. 11/605,943 filed Nov. 28, 2006 in the names of Roderick A. Hyde, et al. and titled “Automated Nuclear Power Reactor For Long-Term Operation”, which application is assigned to the assignee of the present application, the entire disclosure of which is hereby incorporated by reference. Referring to FIGS. 4, 8 and 9, there is shown a specific exemplary first embodiment nuclear fission reactor fuel assembly, generally referred to as 250. Exemplary first embodiment fuel assembly 250 comprises a first loading pattern, generally referred to as 260, for developing and modulating neutron flux level (i.e., neutron population) in first embodiment fuel assembly 250. First loading pattern 260 is shown at a predetermined instant in time after neutron ignition by igniter 245 (e.g., 7.5 years after ignition). The terminology “modulating” is defined herein to mean modifying or changing neutron flux level as a function of time, space and/or energy. Modulating neutron flux level manages reactivity in first embodiment fuel assembly 250. In this manner, the material composition of a region of the reactor is changed. This results in a change of the level of the effective neutron multiplication factor, keff, which in turn results in a change in flux (modulation). As previously briefly mentioned and as disclosed in more detail presently, first loading pattern 260 generates a deflagration wave or “burnfront” 270 that builds excess reactivity into first embodiment fuel assembly 250. Excess reactivity is developed for several reasons, one reason being that more fuel is bred than is burned. First loading pattern 260 balances this excess reactivity sufficiently behind burnfront 270 (i.e., the space between igniter 245 and burnfront 270) while allowing breeding within and in the vicinity of the front of burnfront 270. Referring to FIG. 10, a first control function, generally referred to as 275, corresponding to first loading pattern 260 is shown in graphical form comprising amount of control rod insertion in first embodiment fuel assembly 250 as a function of distance from igniter 245. As seen in FIG. 10, the y-axis is the percentage amount of control rod inserted (the value is 100% behind burnfront 270 and 0% in front of burnfront 270). The x-axis is the distance from igniter 245 shown in units of meters. In the exemplary embodiment illustrated in FIG. 10, the x-axis has a length of approximately four meters. However, this distance may be any suitable distance, such as four meters. This particular example shows a “limit” case. For example, burnfront 270 moves a distance, “x”, and a control rod is fully inserted. Burnfront 270 then moves another distance “Δx” and another control rod is inserted. The step-wise control function shown is a “binary” case. In practice, the reactor operator may deviate from the step-wise function. For example, the control rod closest to burnfront 270 may be in half-way or 50%. First control function 275 modulates neutron flux at a level responsive to changes observed by a monitoring system, as described hereinbelow. As may be appreciated, enhanced steady state deflagration wave burnfront 270 propagation is established by means of a step-wise function type distribution of control material sufficiently behind burnfront 270. As an example, should the rate of reaction fall below desired levels at the front of burnfront 270, a control function response could be to remove or relocate absorber at the rear of the burnfront 270 such that the fission rate increases. A neutron flux level is obtained and the control function is readjusted to again maintain the desired condition. By moving the step function absorber closer to the front of burnfront 270 such that a burn region fission rate is reduced the power can also be reduced. In accident scenarios, it is conceivable to deviate from the step function configurations by placing sufficient absorber throughout the burn wave region. Referring again to FIGS. 8, 9 and 10, first loading pattern 260 comprises control rods 192/194 arranged behind burnfront 270 and centered about a horizontal axis 247b, as shown. First loading pattern 260 further comprises fuel rods arranged into two groups. A first group of fuel rods 280 includes fissile material (referred to herein as a “burning region”) and is arranged in a predetermined first group fuel rod pattern behind burnfront 270 and centered about axis 247b, as described hereinbelow. The burning region is primarily fertile material with some percentage of fissile material bred into it. A second group of fuel rods 290 includes fertile fuel material and is arranged in a predetermined second group fuel rod pattern in front of burnfront 270 and centered about axis 247b, as shown. The terminology “in front of burnfront 270” is defined to mean the space between propagating burnfront 270 and an end of enclosure 155. The terminology “behind burnfront 270” is defined to mean the space between igniter 245 and burnfront 270. Still referring to FIGS. 8, 9 and 10, when igniter 245 releases its neutrons to cause “ignition”, and by way of example only and not by way of limitation, two burnfronts 270 travel radially outwardly from igniter 245 toward ends of enclosure 155, so as to form an oppositely propagating wave-pair. As this occurs, burnfront 270 builds excess reactivity into first embodiment fuel assembly 250 as burnfront 270 propagates from igniter 245 and into first group of fuel rods 280, which are essentially depleted of fissile fuel material. This tends to leave some excess reactivity behind burnfront 270. This result is undesirable because excess reactivity causes increased neutron fluence seen by fuel assembly structural materials in a region behind burnfront 270 where significant burnup has already occurred. Referring again to FIGS. 8, 9 and 10, it should be understood that neutron flux generated by first group of fuel rods 280 behind burnfront 270 breeds fissile fuel material in second group of fuel rods 290 in front of burnfront 270 by transmuting the fertile fuel material in second group of fuel rods 290 into fissile fuel material at the burnfront's leading-edge. Transmuting the fertile fuel material in second group of fuel rods 290 into fissile fuel material at the burnfront's leading-edge advances burnfront 270 in the direction of arrows 295. As burnfront 270 sweeps over a given mass of fuel, fissile isotopes are continually generated as long as neutrons are present to undergo radiative capture in fertile nuclei. The rate at which fissile isotopes are generated may, for a given time and location within the reactor, exceed that of consumption of the fissile isotopes due to parasitic capture and fission. Additionally, capture of neutrons in fertile material leads to intermediate isotopes which decay with a given half-life to fissile material. Because the wave has a propagation velocity, some amount of decay of intermediate isotopes, therefore, occurs behind burnfront 270. A combination of these effects results in additional reactivity remaining and being generated behind burnfront 270. Thus, as shown in FIGS. 8, 9 and 10, burnfront 270 can be modulated to enable a variable nuclear fission fuel burnup. In this type of control configuration, the propagation rate is enhanced by maintaining absorbers as far behind burnfront 270 as allowable to maintain power at a constant level. Biasing of the absorber material behind burnfront 270 counteracts the build-up of excess reactivity within burnfront 270 without reducing the amount of neutrons available for breeding in and in front of burnfront 270. Thus, in order to propagate burnfront 270 in first embodiment fuel assembly 250, burnfront 270 is initiated by igniter 245, as described above and then allowed to propagate. In one embodiment, the actively controllable control rods 192/194 insert neutron absorbers, such as without limitation, Li6, B10, or Gd, into first group of fuel rods 280 behind burnfront 270. Such an insertion of neutron absorbers drives down or lowers neutronic reactivity of first group of fuel rods 280 that is presently being burned by burnfront 270 relative to neutronic reactivity of second group of fuel rods 290 ahead of burnfront 270, giving the wave a propagation direction indicated by arrow 295. Controlling reactivity in this manner increases the propagation rate of burnfront 270 and therefore provides a means to control burn-up above a minimum value needed for propagation and a greater value set by, in part, structural limitations discussed above. Referring to FIG. 11 there is shown an exemplary second embodiment nuclear fission reactor fuel assembly, generally referred to as 300. Exemplary second embodiment fuel assembly 300 comprises a second loading pattern, generally referred to as 310, for modulating neutron flux level in second embodiment fuel assembly 300. Second loading pattern 310 is shown at a predetermined instant in time after neutron ignition by igniter 245 (e.g., 7.5 years after ignition). Modulating neutron flux level manages reactivity in second embodiment fuel assembly 300. As disclosed in more detail presently, second loading pattern 310 generates deflagration wave or “burnfront” 270 that builds excess reactivity into second embodiment fuel assembly 300. Second loading pattern 310 balances this excess reactivity sufficiently in front of burnfront 270 while reducing neutron fluence seen by fuel assembly. Control rods 192/194 insert neutron absorbers into second group of fuel rods 290 in front of the burnfront 270, thereby slowing down the propagation of burnfront 270. In this case, fuel to the left of burnfront 270 is allowed to produce power as the burnfront propagates. One can see that such a control method could lead to the ignition of the entire fuel assembly 300. Referring to FIG. 12, a second control function, generally referred to as 320, corresponding to second loading pattern 310 is shown in graphical form comprising amount of control rod insertion in second embodiment fuel assembly 300 as a function of distance from igniter 245. Second control function 320 modulates neutron flux at a level responsive to changes observed by a monitoring system, as described hereinbelow. Hence, increased steady state deflagration wave burnfront 270 propagation in this embodiment is established by means of a step-wise function type distribution, as shown and is in part dependant on the rate of removal of control rods 192/194. Referring to FIG. 13 there is shown an exemplary third embodiment nuclear fission reactor fuel assembly, generally referred to as 330. Exemplary third embodiment fuel assembly 330 comprises a third loading pattern, generally referred to as 340, for modulating neutron flux level in third embodiment fuel assembly 330. Third loading pattern 340 is shown at a predetermined instant in time after neutron ignition by igniter 245 (e.g., 7.5 years after ignition). Modulating neutron flux level manages reactivity in third embodiment fuel assembly 330. As disclosed in more detail presently, third loading pattern 340 generates deflagration wave burnfront 270 that builds excess reactivity into third embodiment fuel assembly 330. Third loading pattern 340 balances this excess reactivity sufficiently near burnfront 270 (i.e., the space within or adjacent to burnfront 270) via control rods 192/194 which insert neutron absorbers into first group of fuel rods 280 within or to the side of burnfront 270. By allowing build-up and/or utilization of excess reactivity at or near the perimeter of the burnfront, the effective size and velocity of burnfront 270 may be modified. Referring to FIG. 14, a third control function, generally referred to as 350, corresponding to third loading pattern 340 is shown in graphical form comprising amount of control rod insertion in third embodiment fuel assembly 300 as a function of distance from igniter 245. Third control function 350 modulates neutron flux at a level responsive to changes observed by a monitoring system, as described hereinbelow. Steady state deflagration wave burnfront 270 propagation is established by means of a continuous function type distribution, as shown. Referring to FIG. 15, there is shown an exemplary fourth embodiment nuclear fission reactor fuel assembly, generally referred to as 360. Exemplary fourth embodiment fuel assembly 360 comprises a fourth loading pattern, generally referred to as 370, for modulating neutron flux level in fourth embodiment fuel assembly 360. Fourth loading pattern 370 is shown at a predetermined instant in time after neutron ignition by igniter 245 (e.g., 7.5 years after ignition). Modulating neutron flux level manages reactivity in fourth embodiment fuel assembly 360. As disclosed in more detail presently, fourth loading pattern 370 generates deflagration wave burnfront 270 that builds excess reactivity into fourth embodiment fuel assembly 360. Fourth loading pattern 370 balances this excess reactivity sufficiently behind and in front of burnfront 370 through use of control rods 192/194. Loading pattern 370 thereby gives an additional means to control wave size, propagation characteristics, and therefore burn-up and fluence. Alternatively, burnfront 270 may be stimulated “out front” by control rods 192/194 having fissile material therein. Referring to FIG. 16, a fourth control function, generally referred to as 380, corresponding to fourth loading pattern 370 is shown in graphical form comprising amount of control rod insertion in fourth embodiment fuel assembly 360 as a function of distance from igniter 245. Fourth control function 380 modulates neutron flux at a level responsive to changes observed by a monitoring system, as described hereinbelow. Steady state deflagration wave burnfront 270 propagation is established by means of a function type distribution, as shown. Referring to FIG. 17, there is shown an exemplary fifth embodiment nuclear fission reactor fuel assembly, generally referred to as 390. Exemplary fifth embodiment fuel assembly 390 comprises a fifth loading pattern, generally referred to as 400, for modulating neutron flux level in fifth embodiment fuel assembly 390. Fifth loading pattern 400 includes reflector rods 220 in addition to fuel rods 150 and control rods 192/194. As can be seen, by way of a non-limiting example, there is a repeating pattern of a row of absorber with a row of reflector behind the row of absorber. Alternatively, the row of reflector may be located in front of the row of absorber. The reflector returns a portion of the leakage neutrons back towards the absorbing row (and burnfront 270) resulting in a need for less absorber and more neutrons in the burn/breed region. Fifth loading pattern 400 is shown at a predetermined instant in time after neutron ignition by igniter 245 (e.g., 7.5 years after ignition). Modulating neutron flux level manages reactivity in fifth embodiment fuel assembly 390. As disclosed in more detail presently, fifth loading pattern 400 generates deflagration wave burnfront 270 that builds excess reactivity into fifth embodiment fuel assembly 390. Fifth loading pattern 400 balances this excess reactivity sufficiently behind burnfront 370 while reducing neutron fluence seen by fuel assembly materials behind the burnfront as a result of relatively high burnup. Control rods 192/194 and reflector rods 220 modulate neutron flux in the first group of fuel rods 280 behind burnfront 270, thereby changing the effective size and propagation characteristics of the burnfront 270. Referring to FIG. 18, a fifth control function, generally referred to as 410, corresponding to fifth loading pattern 400 is shown in graphical form comprising amount of control rod insertion in fifth embodiment fuel assembly 390 as a function of distance from igniter 245. Fifth control function 410 modulates neutron flux at a level responsive to changes observed by a monitoring system, as described hereinbelow. Steady state deflagration wave burnfront 270 propagation is established by means of a step function type distribution, as shown. As with the embodiment shown in FIGS. 10 and 11, and as described above, this type of distribution leads to an enhanced burnfront propagation rate allowing for a reduced burn-up to achieved. Referring to FIG. 19, there is shown an exemplary sixth embodiment nuclear fission reactor fuel assembly, generally referred to as 420. Exemplary sixth embodiment fuel assembly 420 comprises a sixth loading pattern, generally referred to as 430, for modulating neutron flux level in sixth embodiment fuel assembly 420. Sixth loading pattern 430 is obtained at a predetermined instant in time after neutron ignition by igniter 245 (e.g., 7.5 years after ignition). Modulating neutron flux level manages reactivity in sixth embodiment fuel assembly 420. As disclosed in more detail presently, sixth loading pattern 430 generates deflagration wave burnfront 270 that builds excess reactivity into sixth embodiment fuel assembly 420. Sixth loading pattern 430 balances this excess reactivity sufficiently behind burnfront 270 and in front of burnfront 270 while reducing neutron fluence seen by fuel assembly materials as a result of relatively high burnup. Control rods 192/194 insert neutron absorbers into first group of fuel rods 280 behind and in front of burnfront 270, thereby changing the effective size of the burnfront 270. It may be appreciated that there may be other materials present besides absorber material. Referring to FIG. 20, a sixth control function, generally referred to as 440, corresponding to sixth loading pattern 430 is shown in graphical form comprising amount of control rod insertion in sixth embodiment fuel assembly 420 as a function of distance from igniter 245. Sixth control function 440 modulates neutron flux at a level responsive to changes observed by a monitoring system, as described hereinbelow. Increased steady state deflagration wave burnfront 270 propagation is established by means of a continuous function type distribution, as shown. Referring to FIG. 21, there is shown an exemplary seventh embodiment nuclear fission reactor fuel assembly, generally referred to as 450. Exemplary seventh embodiment fuel assembly 450 comprises a seventh loading pattern, generally referred to as 460, for modulating neutron flux level in seventh embodiment fuel assembly 450. Seventh loading pattern 460 is shown at a predetermined instant in time after neutron ignition by igniter 245 (e.g., 7.5 years after ignition). It should be noted that fuel rods 290 may have already been burnt. Modulating neutron flux level manages reactivity in seventh embodiment fuel assembly 450. As disclosed in more detail presently, seventh loading pattern 460 generates deflagration wave burnfront 270 that builds excess reactivity into seventh embodiment fuel assembly 450. Seventh loading pattern 460 balances this excess reactivity sufficiently behind burnfront 370 while reducing neutron fluence seen by fuel assembly materials as a result of relatively high burnup. Placing the control step function appropriately in front of burnfront 270 while adjusting the control reaction at the rear of burnfront 270 may be performed to reverse the direction of burnfront 270 propagation resulting in wave propagation through previously burned rods 290. Control rods 192/194 insert neutron absorbers into first group of fuel rods 280 that are now arranged behind burnfront 270, thereby changing the effective size of the burnfront 270. Referring to FIG. 22, a seventh control function, generally referred to as 470, corresponding to seventh loading pattern 460 is shown in graphical form comprising amount of control rod insertion in seventh embodiment fuel assembly 450 as a function of distance from igniter 245. Seventh control function 470 modulates neutron flux at a level responsive to changes observed by a monitoring system, as described hereinbelow. Increased steady state deflagration wave burnfront 270 propagation is established by means of a step function type distribution, as shown. It may be understood from the teachings hereinabove, that burnfront 270 can be directed as desired according to selected propagation parameters monitored by a monitoring system. For example, propagation parameters can include a propagation direction or orientation of burnfront 270, a propagation rate of burnfront 270, power demand parameters such heat generation density, cross-sectional dimensions of a burning region through which burnfront 270 is to the propagate (such as an axial or lateral dimension of the burning region relative to an axis of propagation of the burnfront 270), or the like. As another example, the propagation parameters may be selected so as to control the spatial or temporal location, profile and distribution of the burnfront 270, in order to avoid possible failed or malfunctioning control elements (e.g., neutron modifying structures or thermostats), failed or malfunctioning fuel rods, or the like. Failed or malfunctioning fuel rods may be due to swelling or cladding hot spots caused by coolant channel flow blockage. As another example, any ruptured broken fuel rod may be detected by means of feedback provided by detecting tracer isotopes placed within the fuel rods during manufacturing. As a further example, the propagation parameters may be selected based on monitoring or sensing actinides by means of a gas monitor or by sensing of gamma radiation by means of a gamma radiation detector or “Geiger Counter”. As another example, the propagation parameters may be selected based on monitoring data from “coupons” responsive to neutron flux. As yet another example, the propagation parameters may be selected based on measurements of local temperature via thermocouples and flux via neutron detectors. Referring to FIG. 23, a graph illustrates a linear relationship between deflagration wave burnfront velocity and burnup percent versus degree of wave control function. As determined through neutronic simulation, position “A” on the graph corresponds to a step function type of control of burnfront 270 while position “B” on the graph corresponds to distributed type control rod arrangement of burnfront 270. Position “A” corresponds to a configuration similar to that illustrated in FIGS. 9 and 10, while Position “B” corresponds to a configuration similar to that shown in FIGS. 13 and 14. Position “C” on the graph corresponds to a control configuration where absorber is distributed between that of a step function as shown in FIGS. 9 and 10 and that of the continuous function shown in FIGS. 13 and 14; i.e., the absorber is distributed more behind the burnfront than in the distributed case, but not as much as in the step function case. FIG. 23 relates to neutronic results obtained using the MCNPX-CINDER computer software code. In this regard, FIG. 23 shows that if absorber is used, placing the absorber into the reactor as a step function behind the wave gives the fastest wave velocity and the lowest burn-up. Deviation from this configuration (distribute absorber throughout the wave) slows the wave and finally, if absorber is put in front of the wave, the wave's velocity should cease. Referring to FIG. 23A, there is shown a graph illustrating an exemplary spatial distribution of neutron flux, generally referred to as 475. In this regard, the graph plots spatial distribution 475 as neutron flux versus distance from igniter 245. Spatial distribution 475 is representative of the burnfront according to an exemplary control function. Referring to FIG. 23B, there is shown a graph illustrating a spatial profile or control function, generally referred to as 477. Control function 477 corresponds to spatial distribution 475 shown in FIG. 23A. FIG. 23B plots percent of control rod insertion versus distance from igniter 245. Referring to FIG. 23C, there is shown a graph illustrating an exemplary spatial distribution of neutron flux, generally referred to as 479. In this regard, the graph plots spatial distribution 479 as neutron flux versus distance from igniter 245. Spatial distribution 479 is representative of the burnfront according to an exemplary control function. Referring to FIG. 23D, there is shown a graph illustrating a spatial profile or control function, generally referred to as 481. Control function 481 corresponds to spatial distribution 479 shown in FIG. 23C. This graph plots percent of control rod insertion versus distance from igniter 245. Referring to FIG. 23E, there is shown a graph illustrating an exemplary spatial distribution of neutron flux, generally referred to as 483. In this regard, the graph plots spatial distribution 483 as neutron flux versus distance from igniter 245. Spatial distribution 483 is representative of the burnfront. Referring to FIG. 23F, there is shown a graph illustrating a spatial profile or control function, generally referred to as 485, corresponding to spatial distribution 483 shown in FIG. 23E. Spatial profile 485 has a steepest portion 487. This graph plots percent of control rod insertion versus distance from igniter 245. It may be appreciated from the disclosure hereinabove that a burnup value at or below a predetermined burnup value is achievable. In this regard, an amount of neutron absorber, reflector and/or emitter can be controlled at a plurality of locations relative to burnfront 270, such that a majority of the neutron absorption occurs at locations behind burnfront 270 in order to obtain a burnup value at or below a predetermined burnup value. For example, the neutron emitter can be moved from a first location behind burnfront 270 to a second location in front of burnfront 270 to achieve a desired burnup value at or below a predetermined burnup value. In addition, it may be appreciated from the disclosure hereinabove that radiation damage to one or more structural materials can also be controlled in response to controlling burnup in generic fuel assembly 30 and exemplary embodiment fuel assemblies 250/300/330/360/390/420/450. In this regard, controlling such radiation damage would entail achieving a desired radiation damage value, such as DPA, at or below a predetermined radiation damage value. Achieving a radiation damage value at or below a predetermined radiation damage value may comprise moving a neutron emitter from a first location behind burnfront 270 to a second location behind burnfront 270. Alternatively, the neutron emitter can be moved from a first location behind burnfront 270 to a second location in front of burnfront 270 to control potential radiation damage. As another alternative, an amount of neutron absorber can be controlled by means of control rods 192/194 at a location behind burnfront 270 to control potential radiation damage. In this regard, a majority of the neutron absorption due to the neutron absorber may occur at locations behind burnfront 270. In addition, achieving a desired radiation damage value at or below a predetermined radiation damage value may be obtained by controlling an amount of a neutron reflector at a location behind burnfront 270. In this regard, a majority of the neutron reflection due to the neutron reflector may occur at locations behind burnfront 270. It may also be appreciated from the disclosure hereinabove that the neutron flux may be selectively modulated at a location relative to burnfront 270. In this regard, the neutron flux may be modulated at a location behind burnfront 270. In this case, a majority of the modulation occurs at a plurality of locations behind burnfront 270. In addition, selectively modulating neutron flux emitted by burnfront 270 can entail selectively absorbing a portion of the neutron flux emitted by burnfront 270. In other words, an amount of neutron absorber is controlled at a location relative to burnfront 270. More generally, an amount of a neutron interactive material (e.g., insertion of control rods 192/194) can be controlled at the location relative to burnfront 270. In one embodiment, controlling the amount of neutron interactive material at the location relative to burnfront 270 comprises controlling an amount of neutron emitter at the location relative to burnfront 270. The neutron emitter can be a fissile element, a fertile element and/or an element capable of undergoing beta decay to a fissile element. On the other hand, controlling the amount of the neutron interactive material at a location relative to burnfront 270 may comprise controlling an amount of a neutron reflector at the location relative to burnfront 270. In addition, it may be appreciated from the disclosure hereinabove that selectively modulating the neutron flux may be governed by a spatial profile. The spatial profile can be either symmetric or asymmetric with respect to burnfront 270. The spatial profile can have a slope having a steepest portion, the steepest portion suitably occurring at a location behind burnfront 270. It may be further appreciated from the disclosure hereinabove that selectively modulating neutron flux emitted by burnfront 270 may comprise detecting a burning parameter associated with burnfront 270 and selectively modulating the neutron flux at least partially in response to detecting the burning parameter. Detecting the burning parameter may comprise monitoring material radiation damage, such as DPA, at least one location proximate burnfront 270; monitoring burnup at least one location proximate burnfront 270; monitoring burnup speed; monitoring burnfront breadth; monitoring one or more characteristics associated with the neutron flux at least one location proximate burnfront 270; monitoring nuclear radiation at least one location proximate burnfront 270; and/or monitoring temperature at least one location thermally proximate burnfront 270. Moreover, selectively modulating the neutron flux at least partially in response to detecting the burning parameter may comprise selectively modulating the neutron flux at least partially in response to detecting the burning parameter and detecting the burning parameter at least partially in response to a feedback control process and/or at least partially in response to a computer-based algorithm having a plurality of parameters associated with the burning parameter. In this regard, one or more of the parameters of the computer-based algorithm may be modified in response to detecting the burning parameter. Illustrative Methods Illustrative methods associated with exemplary embodiments for controlling burnup in a traveling wave nuclear fission reactor and fuel assembly will now be described. Referring to FIGS. 24-65, illustrative methods are provided for controlling burnup in a traveling wave nuclear fission reactor and fuel assembly capable of emitting a neutron flux. Turning now to FIG. 24, an illustrative method 490 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 500. At a block 510, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor. The method 490 stops at a block 520. Referring to FIG. 25, an illustrative method 530 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 540. At a block 550, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The method 530 stops at a block 560. Referring to FIG. 26, an illustrative method 570 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 580. At a block 590, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 600, a predetermined burnup value is achieved. The method 570 stops at a block 610. Referring to FIG. 27, an illustrative method 620 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 630. At a block 640, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 650, a desired burnup value at or below a predetermined burnup value is achieved. The method 620 stops at a block 660. Referring to FIG. 28, an illustrative method 670 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 680. At a block 690, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 700, the method comprises achieving a burnup value at or below a predetermined burnup value. At a block 710, an amount of a neutron absorber is controlled at a location behind the burnfront. The method 670 stops at a block 720. Referring to FIG. 29, an illustrative method 790 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 800. At a block 810, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 820, a burnup value is achieved at or below a predetermined burnup value. At a block 830, an amount of a neutron absorber achieving neutron absorption is controlled at a plurality of locations relative to the burnfront and wherein a majority of the neutron absorption due to the neutron absorber is at a plurality of locations behind the burnfront. The method stops at a block 840. Referring to FIG. 30, an illustrative method 850 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 860. At a block 870, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 880, a burnup value is achieved at or below a predetermined burnup value. At a block 890, an amount of a neutron reflector is controlled at a location behind the burnfront. The method stops at a block 900. Referring to FIG. 31, an illustrative method 910 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 920. At a block 930, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 940, a burnup value is achieved at or below a predetermined burnup value. At a block 950, an amount of a neutron reflector achieving neutron reflection is controlled at one or more locations relative to the burnfront and wherein a majority of the neutron reflection due to the neutron reflector is at a plurality of locations behind the burnfront. The method stops at a block 960. Referring to FIG. 32, an illustrative method 970 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 980. At a block 990, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 1000, a burnup value is achieved at or below a predetermined burnup value. At a block 1010, a neutron emitter is moved from a first location behind the burnfront to a second location behind the burnfront. The method stops at a block 1020. Referring to FIG. 33, an illustrative method 1030 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1040. At a block 1050, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 1060, a burnup value is achieved at or below a predetermined burnup value. At a block 1070, a neutron emitter is moved from a first location behind the burnfront to a second location proximate to the burnfront. The method stops at a block 1080. Referring to FIG. 34, an illustrative method 1090 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1100. At a block 1110, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 1120, a burnup value is achieved at or below a predetermined burnup value. At a block 1130, a neutron emitter is moved from a first location behind the burnfront to a second location in front of the burnfront. The method stops at a block 1140. Referring to FIG. 35, an illustrative method 1150 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1160. At a block 1170, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 1180, radiation damage to one or more structural materials is controlled in response to controlling a burnup value in the traveling wave nuclear fission reactor. The method stops at a block 1190. Referring to FIG. 36, an illustrative method 1200 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1210. At a block 1220, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 1230, radiation damage to one or more structural materials is controlled in response to controlling a burnup value in the traveling wave nuclear fission reactor. At a block 1240, a radiation damage value is achieved. The method stops at a block 1250. Referring to FIG. 37, an illustrative method 1260 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1270. At a block 1280, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 1290, radiation damage to one or more structural materials is controlled in response to controlling a burnup value in the traveling wave nuclear fission reactor. At a block 1300, a radiation damage value is achieved at or below a predetermined radiation damage value. The method stops at a block 1310. Referring to FIG. 38, an illustrative method 1320 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1330. At a block 1340, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 1350, radiation damage to one or more structural materials is controlled in response to controlling a burnup value in the traveling wave nuclear fission reactor. At a block 1360, a radiation damage value is achieved at or below a predetermined radiation damage value. At a block 1370 a neutron emitter is moved from a first location behind the burnfront to a second location behind the burnfront. The method stops at a block 1380. Referring to FIG. 39, an illustrative method 1390 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1400. At a block 1410, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 1420, radiation damage to one or more structural materials is controlled in response to controlling a burnup value in the traveling wave nuclear fission reactor. At a block 1430, a radiation damage value is achieved at or below a predetermined radiation damage value. At a block 1440 a neutron emitter is moved from a first location behind the burnfront to a second location proximate to the burnfront. The method stops at a block 1450. Referring to FIG. 40, an illustrative method 1460 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1470. At a block 1480, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 1490, radiation damage to one or more structural materials is controlled in response to controlling a burnup value in the traveling wave nuclear fission reactor. At a block 1500, a radiation damage value is achieved at or below a predetermined radiation damage value. At a block 1510, a neutron emitter is moved from a first location behind the burnfront to a second location in front of the burnfront. The method stops at a block 1520. Referring to FIG. 41, an illustrative method 1530 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1540. At a block 1550, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 1560, radiation damage to one or more structural materials is controlled in response to controlling a burnup value in the traveling wave nuclear fission reactor. At a block 1570, a radiation damage value is achieved at or below a predetermined radiation damage value. An amount of a neutron absorber is controlled at a location behind the burnfront at a block 1580. The method stops at a block 1590. Referring to FIG. 42, an illustrative method 1600 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1610. At a block 1620, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 1630, radiation damage to one or more structural materials is controlled in response to controlling a burnup value in the traveling wave nuclear fission reactor. A radiation damage value is achieved at or below a predetermined radiation value at a block 1640. At a block 1650, an amount of a neutron absorber is controlled at a plurality of locations relative to the burnfront and wherein a majority of the neutron absorption due to the neutron absorber is at a plurality of locations behind the burnfront. The method stops at a block 1660. Referring to FIG. 43, an illustrative method 1670 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1680. At a block 1690, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 1700, radiation damage to one or more structural materials is controlled in response to controlling a burnup value in the traveling wave nuclear fission reactor. A radiation damage value is achieved at or below a predetermined radiation value at a block 1710. An amount of a neutron reflector is controlled at a location behind the burnfront at a block 1720. The method stops at a block 1730. Referring to FIG. 44, an illustrative method 1740 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1750. At a block 1760, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. At a block 1770, radiation damage to one or more structural materials is controlled in response to controlling a burnup value in the traveling wave nuclear fission reactor. A radiation damage value is achieved at or below a predetermined radiation value at a block 1780. At a block 1790, an amount of a neutron reflector achieving neutron reflection is controlled at a plurality of locations relative to the burnfront and wherein a majority of the neutron reflection due to the neutron reflector is at a plurality of locations behind the burnfront. The method stops at a block 1800. Referring to FIG. 45, an illustrative method 1810 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1820. At a block 1830, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 1840. The method stops at a block 1850. Referring to FIG. 46, an illustrative method 1860 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1870. At a block 1880, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 1890. At a block 1900, the neutron flux is selectively modulated at a location behind the burnfront. The method stops at a block 1910. Referring to FIG. 47, an illustrative method 1920 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1930. At a block 1940, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 1950. At a block 1960, the neutron flux is selectively modulated at a plurality of locations relative to the burnfront and wherein an amount of modulation at the plurality of locations relative to the burnfront is governed by a spatial profile. The method stops at a block 1970. Referring to FIG. 48, an illustrative method 1980 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 1990. At a block 2000, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 2010. At a block 2020, the neutron flux is selectively modulated at a plurality of locations relative to the burnfront, so that a majority of the modulation of neutron flux occurs at a plurality of locations behind the burnfront. The method stops at a block 2030. Referring to FIG. 49, an illustrative method 2040 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 2050. At a block 2060, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 2070. At a block 2080, a portion of the neutron flux is selectively absorbed at a location relative to the burnfront. The method stops at a block 2090. Referring to FIG. 50, an illustrative method 2100 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 2110. At a block 2120, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 2130. At a block 2140, an amount of a neutron absorber at the location relative to the burnfront is controlled. The method stops at a block 2150. Referring to FIG. 51, an illustrative method 2160 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 2170. At a block 2180, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 2190. At a block 2200, an amount of a neutron interactive material is controlled at the location relative to the burnfront. The method stops at a block 2210. Referring to FIG. 52, an illustrative method 2220 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 2230. At a block 2240, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 2250. At a block 2260, an amount of a neutron interactive material is controlled at the location relative to the burnfront. At a block 2270, an amount of a neutron emitter is controlled at the location relative to the burnfront. The method stops at a block 2280. Referring to FIG. 53, an illustrative method 2290 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 2300. At a block 2310, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 2320. At a block 2330, an amount of a neutron interactive material is controlled at the location relative to the burnfront. At a block 2340, an amount of a neutron reflector is controlled at the location relative to the burnfront. The method stops at a block 2350. Referring to FIG. 54, an illustrative method 2360 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at block 2370. At a block 2380, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 2390. At a block 2400, a burning parameter associated with the burnfront is detected. At a block 2410, the neutron flux is selectively modulated at least partially in response to detecting the burning parameter associated with the burnfront. The method stops at a block 2420. Referring to FIG. 55, an illustrative method 2430 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 2440. At a block 2450, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 2460. At a block 2470, a burning parameter associated with the burnfront is detected. At a block 2480, radiation damage to a material at least one location proximate to the burnfront is monitored. At a block 2490, the neutron flux is selectively modulated at least partially in response to detecting the burning parameter associated with the burnfront. The method stops at a block 2500. Referring to FIG. 56, an illustrative method 2510 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 2520. At a block 2530, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 2540. At a block 2550, a burning parameter associated with the burnfront is detected. At a block 2560, a burnup value is monitored at least one location proximate to the burnfront. At a block 2570, the neutron flux is selectively modulated at least partially in response to detecting the burning parameter associated with the burnfront. The method stops at a block 2580. Referring to FIG. 57, an illustrative method 2590 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 2600. At a block 2610, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 2620. At a block 2630, a burning parameter associated with the burnfront is detected. At a block 2640, burnup velocity is monitored. At a block 2650, the neutron flux is selectively modulated at least partially in response to detecting the burning parameter associated with the burnfront. The method stops at a block 2660. Referring to FIG. 58, an illustrative method 2670 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 2680. At a block 2690, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 2700. At a block 2710, a burning parameter associated with the burnfront is detected. At a block 2720, burnfront breadth is monitored. At a block 2730, the neutron flux is selectively modulated at least partially in response to detecting the burning parameter associated with the burnfront. The method stops at a block 2740. Referring to FIG. 59, an illustrative method 2750 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 2760. At a block 2770, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 2780. At a block 2790, a burning parameter associated with the burnfront is detected. At a block 2800, one or more characteristics associated with the neutron flux are monitored at least one location proximate to the burnfront. At a block 2810, the neutron flux is selectively modulated at least partially in response to detecting the burning parameter associated with the burnfront. The method stops at a block 2820. Referring to FIG. 60, an illustrative method 2830 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 2840. At a block 2850, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 2860. At a block 2870, a burning parameter associated with the burnfront is detected. At a block 2880, nuclear radiation is monitored at least one location proximate to the burnfront. At a block 2890, the neutron flux is selectively modulated at least partially in response to detecting the burning parameter associated with the burnfront. The method stops at a block 2900. Referring to FIG. 61, an illustrative method 3000 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 3010. At a block 3020, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 3030. At a block 3040, a burning parameter associated with the burnfront is detected. At a block 3050, temperature is monitored at least one location proximate to the burnfront. At a block 3060, the neutron flux is selectively modulated at least partially in response to detecting the burning parameter associated with the burnfront. The method stops at a block 3070. Referring to FIG. 62, an illustrative method 3080 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 3090. At a block 3100, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 3110. At a block 3120, a burning parameter associated with the burnfront is detected. At a block 3130, neutron flux is selectively modulated at least partially in response to detecting the burning parameter associated with the burnfront. At a block 3140, the neutron flux is selectively modulated at least partially in response to detecting the burning parameter in response to a feedback control process. The method stops at a block 3150. Referring to FIG. 63, an illustrative method 3160 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 3170. At a block 3180, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 3190. At a block 3200, a burning parameter associated with the burnfront is detected. At a block 3210, the neutron flux is selectively modulated at least partially in response to detecting the burning parameter associated with the burnfront. At a block 3220, the neutron flux is selectively modulated at least partially in response to detecting the burning parameter in response to a computer-based algorithm. The method stops at a block 3230. Referring to FIG. 64, an illustrative method 3240 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 3250. At a block 3260, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 3270. At a block 3280, a burning parameter associated with the burnfront is detected. At a block 3290, the neutron flux is selectively modulated at least partially in response to detecting the burning parameter associated with the burnfront. At a block 3300, the neutron flux is selectively modulated at least partially in response to detecting the burning parameter in response to a computer-based algorithm, wherein the computer-based algorithm incorporates a plurality of parameters. At a block 3310, one or more of the plurality of parameters of the computer-based algorithm is modified in response to detecting the burning parameter. The method stops at a block 3320. Referring to FIG. 65, an illustrative method 3330 for controlling burnup in a traveling wave nuclear fission reactor capable of emitting a neutron flux starts at a block 3340. At a block 3350, the method comprises modulating the neutron flux emitted by the traveling wave nuclear fission reactor, the neutron flux defining a burnfront. The neutron flux is selectively modulated at a location relative to the burnfront at a block 3360. At a block 3370, a burning parameter associated with the burnfront is detected. At a block 3380, radiation damage to one or more of a plurality of structural materials is controlled in response to selectively modulating the neutron flux. The method stops at a block 3390. One skilled in the art will recognize that the herein described components (e.g., operations), devices, objects, and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are contemplated. Consequently, as used herein, the specific exemplars set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific exemplar is intended to be representative of its class, and the non-inclusion of specific components (e.g., operations), devices, and objects should not be taken as limiting. Moreover, those skilled in the art will appreciate that the foregoing specific exemplary processes and/or devices and/or technologies are representative of more general processes and/or devices and/or technologies taught elsewhere herein, such as in the claims filed herewith and/or elsewhere in the present application. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to claims containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that typically a disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms unless context dictates otherwise. For example, the phrase “A or B” will be typically understood to include the possibilities of “A” or “B” or “A and B.” With respect to the appended claims, those skilled in the art will appreciate that recited operations therein may generally be performed in any order. Also, although various operational flows are presented in a sequence(s), it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. Furthermore, terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. Therefore, what are provided are a traveling wave nuclear fission reactor, fuel assembly, and a method of controlling burnup therein. While various aspects and embodiments have been disclosed herein, other aspects and embodiments will be apparent to those skilled in the art. For example, each of the embodiments of the nuclear fission reactor fuel assembly may be disposed in a thermal neutron reactor, a fast neutron reactor, a neutron breeder reactor, a fast neutron breeder reactor, as well as the previously mentioned traveling wave reactor. Thus, each of the embodiments of the fuel assembly is versatile enough to be beneficially used in various nuclear reactor designs. Moreover, the various aspects and embodiments disclosed herein are for purposes of illustration and are not intended to be limiting, with the true scope and spirit being indicated by the following claims.
056446080	claims	1. A method for cooling water from a spent fuel pool of a nuclear power generating plant comprising withdrawing water to be cooled from the spent fuel pool and passing the water to be cooled to and through a heat exchanger for transferring heat from the water to be cooled flowing on one side of a heat exchanger surface of the heat exchanger to a coolant medium flowing on the other side of the heat exchanger surface, comprising causing a flow of air as coolant medium to pass through said heat exchanger while spraying a fine mist of droplets of water into the flow of air to enhance the cooling capacity of the air. 2. The method of claim 1 including causing water from said fine mist of droplets of water to collect on said heat exchange surface as a thin film and evaporating water from said thin film to enhance overall heat transfer. 3. The method of claim 2 and including evaporating at least about one-third of the water sprayed into the heat exchanger within said heat exchanger. 4. The method of claim 1 including causing the flow of air by sucking ambient air through the heat exchanger with a fan. 5. The method of claim 1 comprising spraying the fine mist of droplets into the flow of air through a plurality of spaced nozzles. 6. The method of claim 5 including controlling the spraying of the fine mist by selectively activating only some nozzles of said plurality of nozzles. 7. The method of claim 1 wherein said droplets of water have a mean diameter of about 250 microns or less. 8. The method of claim 1 wherein said droplets of water have a mean diameter of about 150 microns or less. 9. The method of claim 1 wherein said droplets of water have a mean diameter of about 50 microns.
052710521	description	DESCRIPTION OF PREFERRED EMBODIMENT The present inventive system may be employed in any nuclear reactor design which utilizes a boron based chemical shim to control the reactor output. The preferred nuclear reactor system is designed upon a pressurized water nuclear reactor plant configuration since these plant configurations are more apt to employ a soluble neutron capturing material dissolved in the control fluid. Referring to the nuclear reactor plant flow diagram shown in FIG. 1, the process steps for operating a pressurized water reactor utilizing enriched B-10 in the reactor coolant system (RCS) are disclosed. The reactor coolant system consists of the equipment necessary to transport the coolant to and from the reactor core as is known in the art. To the far left of the flow diagram, in block 2, the process starts with a new reactor cycle. The new cycle is characterized by having supplied the reactor vessel with new fuel rods. At this stage, the necessary level of the B-10 boron isotope is found in the reactor coolant system. This B-10 boron level must be high enough to control the output from the new fuel rods once the nuclear reaction is allowed to proceed. The reactor coolant system is said to contain an enriched B-10 boric acid solution in that the ratio of the B-10 boron isotope to the B-11 boron isotope is greater than that of naturally occurring boric acid solutions in which the maximum B-10:B-11 isotope ratio is no greater than 19.8:80.2. The reactor coolant system may also contain other neutron capturing materials, however the preferred reactor coolant system would only contain the EBA solution as the neutron capturing material. Once the nuclear reaction is allowed to proceed, the reactor coolant system solution is circulated through the reactor vessel to control the reaction and also to transfer heat energy to the steam generation system (not shown in diagram). After the reactor fuel begins to be depleted of fissionable material, the level of neutron capturing material in the reactor coolant system solution is decreased so that the maximum output of controllable energy is obtained from the reactor. A reactor coolant system storage loop is provided which is employed to control the concentration of B-10 boron material in the reactor coolant solution. As shown in block 4, a portion of the coolant solution, containing the EBA solution, which is circulating through the reactor vessel is extracted from the coolant system. The extracted EBA solution is selectively directed to either a storage system or a concentration system. As shown in block 6, the EBA (more importantly, the B-10 material or molecules) is preferably stored in ion exchange beds. The ion exchange beds contain an amount of basic anion exchange resin material which can either store the EBA solution or release the EBA solution based upon the temperature of the incoming EBA stream. Thus, the EBA solution can be, in effect, "banked" as the cycle progresses and excess core reactivity decreases. Alternatively, the extracted EBA solution can be directed to the concentration system as shown in block 8. The preferred concentration system is an evaporation system. The extracted EBA solution can be reconcentrated by use of the evaporation system and stored for later use when a reconcentrated EBA solution is required by the reactor system. Both the ion exchange system and the evaporation system can be arranged to return a controlled amount of the EBA solution back to the reactor coolant solution circulating through the reactor vessel. The returned coolant solution may be concentrated to any desired level of B-10 boric acid however the level of B-10:B-11 enrichment is not increased. The solution exiting the ion exchange beds, block 6, may be returned to the reactor coolant system as shown in block 10, or it may be directed towards the evaporation system in block 8 to be concentrated. The ion exchange beds can also release a high concentration EBA solution if a heated solution, preferably between 70.degree.-150.degree. F., and preferably either water or an EBA solution, is supplied via block 38 to the ion exchanger system, block 6. The upper temperature limit for eluting an enriched boric acid solution from the ion exchange resins is limited by the resin utilized. This concentrated EBA solution can be directed towards a concentrated EBA storage system, shown in block 12. The evaporative system in block 8 can supply a concentrated EBA solution to the concentrated EBA storage system, block 12, or a dilute EBA solution to the dilute EBA storage system shown in block 14. The solutions produced in the concentration system can be circulated back to the reactor coolant system as shown in block 16. When some of the reactor fuel rods are nearly depleted of their fissionable matter, the reactive core cycle is concluded. At this point the level of the B-10 isotope in the reactor coolant which is circulating through the reactor vessel is close to 0-10 ppm. All of the B-10 material at this point has either been stored, or "banked", in the ion exchange resin beds or in the concentrated EBA storage facility or have been lost due to reaction with a neutron particle. The concentration of the depleted reactor coolant solution is shown in block 18. Once the reactor has reached the end of its cycle, some of the reactor fuel rods must be replaced. First, the reactor coolant system is "borated" as shown in block 19. This step is accomplished by a "feed and bleed" procedure in which an NBA solution is fed into the reactor coolant and the depleted reactor coolant in the reactor coolant system is removed. Next, the reactor coolant system is drained to the reactor flange level. The reactor vessel head is now removed and the refueling canal and reactor vessel cavity are flooded with an NBA solution from the refueling water storage tank, as shown in block 20. Once refueling has taken place, an amount of the refueling solution is transferred via the drain system back to the refueling water storage tank and the reactor head is replaced, as is shown in block 22. The solution now residing in the reactor coolant system (including the reactor vessel) must be displaced, preferably to the greatest extent possible, before introducing the EBA solution necessary to control the nuclear reaction of the next cycle. A portion of the solution in the reactor coolant system is sent to the refueling water storage tank or preferably to the reactor hold-up tanks. This step is shown in the figure as block 26 and 28. The lowest level of drainage will depend upon safety factors to ensure that the reactor core is at all times properly cooled and controlled. The solution transferred to the reactor hold-up tank can also be further processed to reconcentrate the natural boric acid. The next step in the process is to displace the residual solution in the reactor coolant system and the vessel with the EBA solution to be used in the reactor coolant system, as is shown in block 30. Upon displacement, the solution exiting the reactor coolant system (and vessel) is directed to the reactor hold-up tank, as is shown in block 32. The EBA solution which is used to displace the residual natural boric acid solution remaining in the reactor coolant system can be supplied by various systems within the reactor plant. First, the EBA solution can be obtained from the concentrated EBA storage system, block 12. An alternative source of the EBA solution is to use the EBA make-up system as is shown in block 36. This EBA make-up system is merely a combination of processing units which can supply a highly concentrated enriched boric acid solution. The entire reactor coolant system is then refilled with a solution of EBA, block 34, using the same EBA sources as for the displacement of the refueling NBA solution. The beginning of the next cycle can proceed as is shown in block 2. The boron-10 to boron-11 isotope ratio in the reactor coolant system may vary from greater than 19.8:80.2 to 95:5 at the start of a reactor cycle. A relatively small amount of make-up EBA will have to be introduced to the reactor coolant system during the reactor cycles. This is due to the fact that the B-10 nuclei reacts during the reactor cycle with neutrons and also since not all of the reactor coolant solution can be drained during refueling and therefore some dilution of the EBA solution in the reactor coolant system occurs when the refueling takes place. Therefore, an EBA make-up supply is necessary as is shown in block 36. In order to more fully describe the steps of the preferred method of displacing the remaining refueling NBA solution in the reactor coolant system by an EBA solution, and minimizing the intermixing of the two solutions, reference is made to FIG. 2. A reactor vessel 90 and reactor vessel head 91 containing a reactor core 25 is shown with the reactor vessel 90 containing a minimum level of solution to control the reaction, shown representatively as fluid level 84. This stage of refueling corresponds to block 26 in FIG. 1. Optional loop stop valves 31, 41 are shown downstream of the reactor coolant system pump 35 and upstream from the steam generation system 40. Prior to the EBA solution switch over, the reactor core temperature is maintained by circulating the solution of NBA within the reactor vessel 90 by residual heat exchanger pump 60 through the residual heat exchanger 50. The valves 51, 52 are open to allow the NBA solution to flow through this residual heat exchange loop. Optional loop stop valves 31, 41 would be closed if the particular plant design incorporated these valves. The EBA refueling valve 53 is open during this NBA core cooling operation. The process step of draining the reactor vessel 90 to its minimum level prior to displacement with the EBA solution, as shown in FIG. 1 by block 26 and 28, is preferably accomplished by opening the reactor hold-up tank value 81 and draining a portion of the solution in the reactor vessel 90 to the reactor hold-up tank 80. In order to displace the NBA refueling solution with an EBA solution, the EBA solution source valve 71 and the reactor hold-up tank valve 81 are opened, and the EBA refueling valve 53 is closed. This diverts the solution exiting the reactor vessel 90 to the reactor hold-up tank 80 while the incoming EBA solution from the EBA solution source 70 fills the reactor vessel 90. The EBA solution source 70 can be flow coupled to either the ion exchange system, the evaporation system, the concentrated EBA storage system, or to the EBA make-up system, however, it is preferred to connect to the concentrated EBA storage system. By employing the use of loop stop valves 31, 41 the amount of NBA solution which enters the reactor coolant system is kept to a minimum. When the EBA solution has displaced the NBA refueling solution the EBA solution source valve 71 and the reactor hold-up tank valve 81 are closed. Normal operating procedures are now followed to bring the reactor back on line for power production. Eventually, when the reactor is prepared to be put back in the power production mode, the RHR system can be shutdown until needed. The valves 51, 52 are closed. Valve 53 may optionally be reopened. The residual heat exchange pump 60 is shut down. If loop stop valves 31, 41 are utilized, they are opened again at this time. The residual heat removal system is thus shut down until needed. Only one residual heat removal loop is shown in FIG. 2, however it is known in the art to supply a plurality of such residual heat removal loops and the present invention can be employed with each residual heat removal loop. The displacement of the NBA refueling solution in the reactor vessel 90 by the EBA replacement solution can be monitored by two methods. First, if the incoming EBA replacement solution is at a different temperature than the NBA refueling solution, the displacement is discontinued when the exiting temperature of the solution directed towards the reactor hold-up tank 80 is approximately that of the replacement EBA solution. The displacement could be discontinued when the temperature difference between the exiting solution and the EBA solution is some predetermined value. Preferably, the temperature difference between the two solutions is maximized to enhance this processing step. Also the boron-10 enrichment of the exiting solution can be monitored and displacement is discontinued when the boron-10 enrichment of the solution directed towards the reactor hold-up tank 80 approaches the level of the replacement EBA solution. The temperature or boron-10 enrichment level are monitored by the analyzer 83. If the temperature is analyzed, analyzer 83 is a temperature sensor. If the boron-10 enrichment level is analyzed, the analyzer 83 is a mass spectrometer. Optionally, both detection systems could be utilized as a design alternative. The temperature and the B-10 enrichment level of the replacement enriched boric acid solution is determined by a similar analyzer device which is not shown. The intermixing of the EBA replacement solution and the NBA refueling solution is dependent upon the plug flow displacement process. The approximate percentage of the NBA solution that would intermix with the EBA solution and be present in the reactor coolant solution after displacement without the use of loop stop valves 31, 41 is approximately about 30%. That is, 30% of the final reactor coolant system volume would be comprised of the NBA solution from the refueling step and about 70% would be the replacement EBA solution. Employing the use of loop stop valves 31, 41 would lead to an after refueling volume of NBA in the reactor coolant system of approximately 10% with remaining 90% being the EBA solution. The operation of a nuclear reactor with an EBA solution in the reactor coolant system allows for maintaining a milder chemistry within the reactor coolant system. The prior art method of reactor operation with a NBA solution required a higher boric acid concentration and therefore a higher concentration of lithium hydroxide for pH balancing purposes. The EBA process of the present invention allows for the same effective B-10 concentration in the reactor coolant with its accompanying control capacity, while decreasing the total boric acid concentration required to supply that level of the B-10 isotope. Due to the higher expense of an EBA solution as opposed to an NBA solution, the present invention provides for procedures which minimize the mixing between the EBA solution in the primary reactor coolant system and the NBA solution in the other boric acid systems. This allows for the operation of the nuclear reactor plant using the EBA solution and therefore taking advantage of the associated benefits while minimizing the added incremental costs of such an operation .
	summary
	description	The utility patent application claims the benefit of U.S. Provisional Patent Application No. 62/565,293, filed on Sep. 29, 2017, the entirety of which is incorporated by reference. The U.S. Government has rights in this invention pursuant to Contract No. DE-AC02-06CH11357 between the U.S. Department of Energy and Chicago Argonne, LLC, representing Argonne National Laboratory. This invention relates to medical isotope production and more specifically, this invention relates to a system and method for producing medical isotopes at efficiencies and production rates that are magnitudes higher than state of the art protocols. Electron linear accelerators or LINACs are used widely through the medical world in various forms for radiation therapy and other applications. The electron beams may be used directly. Alternatively, the electron beams may be converted to gamma beams upon collision with a converter comprising a high atomic number element. These secondary gamma beams are produced by a physical process called Bremsstrahlung (braking radiation) wherein incoming electrons are deflected, or accelerated, by nuclei of atoms in the converter material. The gamma beams can then be used to irradiate targets positioned downstream so as to produce medically important isotopes via photonuclear reactions. Production of gamma beams of sufficient energy for such reactions requires electron linacs with electron beam energies up to about 40 MeV. The electron to gamma conversion process is very inefficient so that high electron beam power is generally needed to produce a high flux of gamma rays which in turn is needed to produce isotopes at a high rate. In many cases, the target materials needed to produce desired isotopes are very rare and/or expensive. Hence, there is also the need to develop geometries for the production assembly that minimizes both the required beam power and the amount of the expensive or rare target material. Medical isotopes which emit alpha particles are especially sought after for cancer treatment. The inherent benefits of alpha particles are that they have a high linear energy transfer (LET) in most materials, thereby delivering nearly all their energy within a very short distance (approximately 40-100 microns). Such isotopes can be incorporated in molecules to target tumors and kill cancer cells, which are in the size range of 2-120 microns. If targeted accurately, this treatment greatly reduces damage to surrounding healthy tissue, compared to methods of radiation treatment that employ either direct electron beams or gamma beams or medical isotopes which decay by beta particle emission. (These other moieties are not of the high linear energy transfer (LET) variety and therefore much more penetrating. High LET radiation is very effective at killing DNA within tumor cells because it induces a large number of double strand breaks due to the high deposited energy density.) Some medical isotopes are hard to make cost-effectively in useful amounts. Most existing methods require a combination of large amounts of isotopically enriched target material, high beam power, more expensive accelerators, and/or processing of large quantities of radioactive material. Hence, most existing solutions are too expensive, especially for certain therapeutic isotopes, such as 225Ac. For example, the U.S. Department of Energy currently extracts less than 1 Ci 225Ac per year from a large amount of radioactive material at a very high cost per mCi. Another method currently under development requires irradiation and chemical processing of large amounts of target material after radiation with high energy proton accelerators. Medical isotopes produced using electron beams presently use electron to gamma converters which are thin plates of materials such as tantalum or tungsten immersed in flowing water for cooling. This limits the beam sizes to large diameters so as to prevent a concentration of energy on a smaller area that would otherwise boil the coolant water. Impingement of the large diameter beams on the converter plates requires that the plates be followed by large diameter targets, upon which impinge gamma particles. The size of these targets require that they contain large amounts of enriched isotopes or otherwise rare materials, sometimes up to 50 to 100 grams. This is expensive. A need exists in the art for a system and method for yielding large amounts of medical isotopes with very small amounts of target material, especially for cases in which the target material is very rare or expensive. The system should define a very small foot print. Preferably, the beam collimator, the electron-to-gamma converter, and the target are integrated into an optimized, compact assembly. To make maximum use of the gamma flux diverging from the converter, the distance between the converter and target material should be minimized. The resulting production assembly then would have a high yield of the desired isotope relative to the input beam power while, also, using a minimal amount of target material. An object of the invention is to provide a system and method for producing medical isotopes that overcomes many of the drawbacks of the prior art. Another object of the invention is to provide a system and method for efficiently producing medical isotopes. A feature of the invention is utilization of both small beam and target diameters such that the beam diameter is as small as the diameter of the target so that the entire area of the target is contacted by the beam. An advantage is that the very high power density results in efficiently producing medical isotopes with minimal total beam power and, also, a minimal amount of target material. The optimal beam and target diameters are determined by Monte Carlo simulations. Preferably, the optimal beam and target diameters are about 2-3 mm. Target diameter is small because of the small mass of the target. In an embodiment of the invention, the beam cross section or diameter is about the same as the cross section or diameter of the target. Alternatively, the beam is relatively smaller in cross section or diameter by ˜10 percent to 30 percent. For example, a 3 mm diameter target may be contacted by a 2 mm diameter beam. The invention also incorporates a method of cooling the collimator and converter that enables delivering the required beam power into the required small diameter spot. In an embodiment of the invention, the converter does not directly contact the coolant. For example the converter is not directly immersed in coolant water. Still another object of the invention is to provide a method for cooling a system for producing medical isotopes, so that the system can utilize high energy densities. A feature of the invention is an optimized cooling configuration whereby the collimator, converter and target container are maintained at or below their respective melting points. The collimator module, converter module and target module are independent but well aligned with each other and spaced closely to form an overall compact system. Water cooling is applied at a distance such that the water does not boil. The materials directly irradiated by both the primary electron and secondary gamma beams are chosen to have the required physical properties such as high melting point and high thermal conductivity. Optionally, the assembly may be operated in an inert atmosphere or high vacuum such that the materials do not corrode via chemical reactions with the surrounding air. An advantage of the invention is that very high power densities can be utilized to produce large amounts of medical isotope from relatively low electron beam power, narrow beam widths, and small amounts of target material. Yet another object of the present invention is to provide a compact device for producing medical isotopes. A feature of the device is its integration of a beam viewer, collimator, an electron-to-gamma converter, and target material into a single, compact nacelle. An advantage of the invention is that it yields large amounts of medical isotopes using photonuclear reactions with relatively low beam power (approximately 10-20 kW) and small amounts of target material (approximately from 50 milligrams to a few grams). Briefly, the invention provides a method for generating medical isotopes, the method comprising contacting a radiation beam with a converter for a time sufficient to produce a beam of gamma particles, and contacting the beam of gamma particles to a target, where the cross section dimension of the beam of gamma particles is similar to the cross section dimension of the target. An example of the method is directing a small diameter electron beam to a converter material to produce a secondary beam of gamma particles which, in turn, impinge upon a down-stream target, where the cross section dimension of the beam of gamma particles is similar to the cross section dimension of the target. Since the gamma beam diverges from the converter material, it is necessary to minimize the distance separating the converter from the target. Also, since a large amount of energy is deposited by the electron beam in the converter material, while a much smaller amount is deposited in the target material (mostly by the gamma flux), these 2 components must not be in thermal contact. Also provided is a system for producing medical isotopes, the system having a first upstream end and a second downstream end; a radiotransparent channel with a first upstream end and a downstream end, wherein the upstream end is adapted to receive a radiation beam; and a target positioned downstream of the downstream end of the channel and coaxially aligned with the channel, wherein the target has a cross section that is similar to the cross section of the channel. The foregoing summary, as well as the following detailed description of certain embodiments of the present invention, will be better understood when read in conjunction with the appended drawings. All numeric values are herein assumed to be modified by the term “about”, whether or not explicitly indicated. The term “about” generally refers to a range of numbers that one of skill in the art would consider equivalent to the recited value (e.g., having the same function or result). In many instances, the terms “about” may include numbers that are rounded to the nearest significant figure. The recitation of numerical ranges by endpoints includes all numbers within that range (e.g. 1 to 5 includes 1, 1.5, 2, 2.75, 3, 3.80, 4, and 5). The following detailed description should be read with reference to the drawings in which similar elements in different drawings are numbered the same. The drawings, which are not necessarily to scale, depict illustrative embodiments and are not intended to limit the scope of the invention. As used herein, an element or step recited in the singular and preceded with the word “a” or “an” should be understood as not excluding plural said elements or steps, unless such exclusion is explicitly stated. As used in this specification and the appended claims, the term “or” is generally employed in its sense including “and/or” unless the content clearly dictates otherwise. Furthermore, references to “one embodiment” of the present invention are not intended to be interpreted as excluding the existence of additional embodiments that also incorporate the recited features. Moreover, unless explicitly stated to the contrary, embodiments “comprising” or “having” an element or a plurality of elements having a particular property may include additional such elements not having that property. The invention provides a method for generating medical isotopes, the method comprising delivering an electron beam from an accelerator to impinge upon a converter to produce a secondary beam of gamma particles. The beam of gamma particles travel on to a target in which photonuclear reactions transmute nuclei of the target to produce atoms of the sought after medical isotope. The target is located as close as possible to the converter so that the diameters of both the gamma beam and target are similar (within 25 percent, preferably within 15 percent, and most preferably within 5 percent) to the diameter of the original electron beam. It is most effect to have the beam diameter no larger than the target diameter. For example, if the target is 100 units in diameter, the beam may vary from 75 to 100 units in diameter, and most preferably from 80 to 100 units. The invention provides high yield, efficient production of medical isotopes using photonuclear reactions. For example, electron beams are used in conjunction with very small targets comprising rare target materials such as 226Ra. The invention's primary purpose is the production of a medical isotope 225Ac and its daughter 213Bi, both of which can be used for cancer therapy. However, the invention is applicable to producing increased levels of other isotopes which heretofore require large amounts of separated isotopes as target material. Also, aside from producing 225Ac, the invented method and system also can economically produce 224Ra/212Pb simultaneously in the same 226Ra target. In general the method has advantages for any photonuclear reactions that require very rare or expensive target materials, such as 48 Ca. Simulations of various isotopes have determined that the system can generate in one irradiation: 225Ac, also 225Ra/225Ac/213Bi as a generator, and a 224Ra/212Pb generator. In summary, final products can be the 3 alpha emitters: 225Ac, 213Bi, and 212Pb. FIG. 1 depicts the invented system, generally designated as numeral 10. The system is an assembly of 3 modules, first a collimator module, second a converter module, and third a target module. The three modules are thermally isolated but arranged in close proximity to each other to maximize yields of the produced isotope. The distance from the collimator to the converter is not critical, but the distance from the converter to the target should be minimized to achieve high yields per unit of target material. Accurate axial alignment of the 3 modules is guaranteed mechanically. FIG. 1A depicts a collimator 12 with a first upstream end 14 and a second downstream end 17. The housing contains virtually all of the components of the invented system and is designed for compactness. As such, the footprint of the housing is approximately 10 cm long by 5 cm wide. The upstream end 14 is adapted to receive an incoming radiation beam 9. This upstream end 14 may define a window 15 defining a frusto-conical shaped channel 16 with a first upstream end 18 and a second downstream end 20. The first end of the channel 16 defines a wider cross section than the second end so as to distribute the outer portion of the impinging electron beam 9 over a larger area of the collimator. FIG. 1B is a cross section of a similar structure, sans the frusto-conical shaped channel (element 16 in FIG. 1A). Generally, FIG. 1B is identical in function with the collimator depicted in FIG. 1A. However, the collimator depicted in FIG. 1B features a secondary vacuum line, 50. This vacuum line is positioned at the downstream end 17 of the collimator. The assembly is connected to the electron beam vacuum pipe on the upstream side. A vacuum is also created in the small space between the converter (tungsten) and a thin titanium “window” on the downstream target side. This small space is pumped via a “closed” vacuum system for radiation safety reasons. In case of a breach of the target window no radiation will be released to the environment. FIG. 1B also features a different means for coupling the precollimator module to the converter module, that means designated as numeral 52, and comprising a semi rigid band such as a “quick disconnect” flange. This quick disconnect feature allows the collimator module to remain in place while the converter/target assembly is removed for off-line processing in a hot cell. By contrast, FIG. 1A comprises a bolt/threaded aperture configuration integrally molded with its interlocking collar 36 for coupling/decoupling the collimator from/to the converter/target assembly. A salient feature of the invention is that the cross section of the target 34 is very similar in size to the cross section portion of the incoming radiation beam 9 which passes through the collimator 12 so as to utilize virtually the entire bulk of a typically very expensive target 34. The channel 16 is encircled by a heat sink material 22 so as to be in physical contact with the material. Exemplary material is a metal selected from the group consisting of copper, Glidcop®, aluminum, and combinations thereof. Glidcop is the registered trademark name of North American Höganäs, that refers to a family of copper-based metal matrix composite (MMC) alloys mixed primarily with aluminum oxide ceramic particles. The addition of small amounts of aluminum oxide has minuscule effects on the performance of the copper at room temperature (such as a small decrease in thermal and electrical conductivity), but greatly increases the copper's resistance to thermal softening and enhances high elevated temperature strength. The addition of aluminum oxide also increases resistance to radiation damage. As such, the alloy has found use in applications where high thermal conductivity or electrical conductivity is required while also maintaining strength at elevated temperatures. Peripheral regions of the heat sink material 22 define channels 24 adapted to receive coolant fluid, such that the channels encircle the beam 9 at a radially displaced position, relative to the beam. The typical cooling fluid is water entering at a temperature ˜25 degrees C. and exiting at a temperature less than 100 degrees C. The heat sink material 22 may be encircled by cladding material 26 such that radial aspects of the channels 24 terminate in medially facing surfaces of the cladding material. In this configuration, coolant coursing through the channels 24 are in simultaneous physical (and therefore thermal contact) with both the heat sink material 22 and the cladding 26. (The cladding material may also be similar to the heat sink material, which is to say copper, aluminum, Glidcop, etc.) The coolant channels may encircle substantially the entire length of the channel 16, or at least enough of it to prevent its melting during system operation. The second downstream end 22 of the channel terminates at an upstream end of a tubular shaped extension 28. The cross section diameter of this tubular shaped extension is similar to the cross section diameter of the target 34 situated further downstream. This tubular shaped extension 28 is similarly encircled by downstream regions of the heat sink material 22. However, these downstream regions of the heat sink material may or may not form coolant channels in their peripheries along the entire length of the tubular extension 28. The highest energy density in the entire assembly is in the converter. The water cooling of the converter module must carry away this energy. The channel 16 and its downstream tubular extension 28 combine to form a continuous radio-transparent conduit through which the beam 9 may travel. The downstream end of the tubular extension 28 terminates in a cavity 30. The beam diameter entering the collimator at its upstream aperture or iris 15 could typically by about 3 mm while the beam exiting will be the diameter of the tubular region 20-28 is typically about 2 mm. Beam collimation ensures that only useful beam enters the converter module. Beam of larger diameter than the collimator aperture is generally not preferred as such a wide beam deposits its energy in the collimator body and this energy is carried away by the water cooling of the collimator. This can be somewhat ameliorated by the frustoconical shape of the beam channel featured in FIG. 1A. Positioned downstream from the cavity 30 is heat sink material 22 similar to the aforementioned heat sink material 22 encircling the channel 16. Portions of this downstream heat sink material define an aperture 32 coaxial with the channel 16, also previously described. The aperture 32 extends completely through the downstream heat sink material 22 and terminates at its downstream end at a point proximal with a converter 33. This downstream module is reversibly and rigidly attached to the upstream module via an interlocking collar 36. Peripheral regions of the collar may define apertures aligned with threaded apertures formed in peripheral regions of the heat sink 22 and/or cladding 26. These apertures are adapted to receive bolts so as to removably attach the collar to the cladding. A salient feature of the invention is that the converter 33 is in physical contact (for example by brazing) with the heat sink material 22, the latter of which is cooled via coolant flowing about the periphery of the heat sink material 22. All but an upstream facing surface of the converter is embedded into the heat sink material 22 so as to maximize thermal conductance between the converter and surrounding cladding material 26. The upstream facing surface of the converter 33 is in fluid communication with the channel, channel extension 28 and the downstream heat sink aperture 32. This configuration results in the converter being isolated from any coolant fluid such that no coolant fluid (such as water) contacts the converter 33. This feature enables operation of the converter at temperatures much higher than the boiling point of water which is a limiting feature of converter assemblies comprising thin plates immersed in flowing water. Optionally, the upstream facing surface of the converter 33 may be bathed in a relatively inert fluid (e.g., a gas such as argon, helium, nitrogen) so as to prevent exposure of the converter to air. The inert fluid (such as a gas) may be supplied via the distal end of an inert fluid conduit 35, that distal end positioned in close spatial relationship to the upstream facing surface of the converter. A proximal end (not shown) of the conduit 35 would be in fluid communication with an inert gas supply (also not shown). The inert gas can exit through the collimator channels 16, 20, 28. Furthermore, the inert fluid may be provided at a temperature below ambient temperature so as to further aid in maintaining the converter at a predetermined temperature (e.g., the converter melting point). In an embodiment, the upstream collimator and converter modules are contained within the electron linac beam vacuum and there is a small vacuum space 54 (FIG. 1B) between the downstream side of the converter and the thin window in front of the target. Alternatively, this small space can contain flowing inert gas within a closed system to ensure no radiation is released as described above for the embodiment using a closed vacuum system. The target 34 is positioned downstream from the converter 33 and may be separated by a thin space so as to not contact the converter. A downstream surface of the target is shown in FIG. 1A as being backstopped and otherwise in physical contact with a frustoconically shaped end cap 38. (The conical shape is one possible method to enable more convenient loading of radioactive target material such as 226Ra via a remote manipulator in a hot cell.) Peripheral regions of the end cap 38 may define apertures in alignment with similarly sized threaded apertures along the periphery of a downstream facing surface of the cladding 26. These apertures are adapted to receive threaded bolts so as to removably attach the cap 38 to the second end 17 of the system. Suitable coolant fluid is that which has a boiling point of at least 100° C., usually water. As such, a myriad of coolants may be utilized, including but not limited to water, polyethylene glycol, other alcohols, subcritical fluids and combinations thereof. Longitudinally extending exterior surface regions of the cladding 26 may define a first means of coolant ingress 42 while a second portion of the cladding 26 may form a first means of coolant egress 44. Suitable means include conduit or piping as depicted in FIG. 1. The coolant may be supplied under pressure and recirculated via a pump after passing through an external heat exchanger. Alternatively, the coolant may be drawn into the housing via a negative pressure applied to the depending end of the coolant egress conduit 24. The means of ingress 42 and/or egress 44 may include a valve 46 to control fluid flow into and out of the housing. Flow rate of the fluid is empirically determined, so as to prevent liquid fluid from vaporizing or appreciably increasing in temperature to the point of creating a pressure breach, or physical compromise to components of the system. The collimator module and the converter module are separately limited in temperature by a coolant at their outer diameters. The target module is similarly cooled. Unlike state of the art systems, only small amounts of isotopically pure target are required. The inventors have determined through detailed simulations that about 100 mg of 226Ra target material is required to produce about 1 Ci of 225Ac, for example, using a 10-kW electron beam through a 2 mm diameter collimator onto the converter during a 5-day irradiation. Energy/Power Density Detail This new geometry overcomes the power density limitations of conventional converter plates immersed in water. Generally, radiation energies of between about 10 kW and about 20 kW are provided. In operation, an incoming electron beam 9 is provided that has just slightly less diameter or cross section as the target 34. This assures that virtually all of the target (e.g. 226Ra) is contacted by the secondary gamma beam. It is this feature that results in a more than 16-fold increase in production of desired medical isotope at the same beam power and target mass as the present art. For example, the beam 9 may have a diameter at the converter of 2 mm and the target 34 located immediately behind the converter may have a cross section diameter of 3 mm. The beam 9 travels through the channel 16, tubular extension 38 and downstream aperture 32 to impinge upon the converter. The secondary gamma beam diverges slightly from the converter to impinge upon the slightly larger target located immediately behind the converter. Upon conversion, the secondary radiation as gamma particles impinge upon the target, inducing nuclear transmutation of some target atoms to the medical isotope of interest. During the entire process, a coolant serves as a heat sink to maintain the components of the system at below the boiling point of the coolant. Specifically, the coolant carries away the energy deposited in the three separate modules 11, 13, 15. The radius of the coolant channels is adjusted in the design simulations to ensure enough surface area to keep the coolant temperature below its boiling point. Suitable coolant is that which maintains system temperature below the melting point of its lowest melting point component. Alpha particle-emitting isotopes are cytotoxic agents for enabling targeted therapy. Properties of alpha particle radiation such as their limited range in tissue of a few cell diameters and their high linear energy transfer leading to dense radiation damage along each alpha track are promising in the treatment of cancer, especially when single cells or clusters of tumor cells are targeted. Actinium-225 (225Ac) is an alpha particle-emitting radionuclide that generates 4 net alpha particle isotopes in a short decay chain finally leading to stable 209Bi. More than a Curie of the alpha emitter 225Ac can be produced using this invention from 100 mg of 226Ra target. The transmutation is initiated with a photo-neutron reaction as depicted in Equation 1, below:226Ra+→225Ra+n Equation 1 A photon of significant energy (>6.4 MeV) liberates a neutron from Ra-226, leaving a Ra-225 atom. Ra-225 then undergoes natural radioactive beta decay (half-life 14.9 days) to Ac-225, as indicated in Equation 2, below:225Ra→225Ac+β Equation 2 The radioactive isotope 225Ra can be accumulated during a 5-day irradiation using only 10 kW of beam power on the converter. With the assembly shown in FIG. 1A or 1B, the 10-kW beam on the converter for about 5 days will produce ˜1 Ci of 225Ra and 0.1 Ci of 225Ac. As 225Ra decays to produce more 225Ac, the 225Ac can be extracted via “milking” of the Ac from Ra via an ion-exchange resin. After several such milkings of ˜100-200 mCi of 225Ac, an integrated yield of >1 Ci of 225Ac is available for therapeutic use (see below). Table 1 below itemizes yields of photonuclear production of 225Ac from 100 mg of 226Ra target. At the end of the 5-day run 1.1 Ci of 225Ra is produced. 225Ra has a half-life of 15 days and decays to 225Ac which has a half-life of 10 days. The table indicates that the 226Ra is irradiated for 5 days. Then after 2 more days on day 7, the total accumulated amount of 225Ac is 292 mCi. Then on the 7th day 292 mCi of 225Ac can be “milked,” i.e. extracted from the 226R a target. Then additional amounts of 225Ac can be milked from the target once every few days due to the continuing decay of 225Ra to 225Ac. Finally, after 8 “milkings” a total of >1 Ci of 225Ac can be extracted from the initial about 1 Ci of 225Ra, as indicated at the bottom of the table. The 225Ac produced is free of 227Ac contamination because the gamma beam can knock a neutron from the 226Ra target, but not cause a neutron to be absorbed to make 227Ra which would decay to 227Ac. Other methods of producing 225Ac can simultaneously produce 227Ac as an impurity. TABLE 1Photonuclear Production of 225AC from 100 mg of 226Ra225AcDay(mci)Start0RunStop5DeliveriesRunMilk7292.01Milk10185.92Milk13161.73Milk17176.84Milk22173.15Milk28155.36Milk33155.21144.9 mCi from a 5-day once per month7Milk48100.88Total1400.91400.9mCi from a 5-day run every 2 months “Milking” is the process of extracting the accumulated quantity of the short-lived daughter product of the decay of the longer-lived initial material, e.g. the 10 day half-life 225Ac from the mother isotope 15-day half-life 225Ra material. This can be done many times until the initial mother isotope has decayed away. FIG. 2 is a plan view of the invented system, taken along lines 2-2 of FIG. 1. In an embodiment of the invention, the collimator module is shown beveled at an angle, typically 45 degrees. The beam halo surrounding the small diameter collimator aperture causes “visible optical transition radiation” which enables viewing of the beam at the edges of the collimator opening in order to ensure the high-power beam is centered on the entrance to the collimator. This visible radiation is viewed via a shielding video camera viewed using a mirror in the beamline upstream of the collimator. FIG. 3 depicts the heat dissipation characteristics of the collimator, converter and target modules of the invented system, along the same view as FIG. 1. Similar simulations have been done for the converter module and the target module. Less heat is generated in the precollimator module 11 of the device, compared to the converter module 13 and target module 15. The highest energy density is deposited in the converter 33. About 4 kW of the 10 kW transmitted through the collimator to the converter are deposited in the converter. The remaining 6 kW represents the power converted to gamma rays. The peak temperature in the converter is about 1000° C. which is well under the melting point of tungsten. At the outer diameter of the tungsten converter, the temperature is well under the melting point of the braze material. Also, the temperature at the larger radius of the aluminum, copper or Glidcop where the cooling channels are located, is under the boiling point of water. This example is based on simulated data. The primary purpose of this simulation is to produce >1 Ci/week of 225Ac (10 day half-life). A co-produced isotope that is also of great interest is 212Pb (10 hour half-life) which will be available at greater than 10 Ci/week. These yields are based on an assumed beam power of 11 kW of 40-MeV electron beam for 5 days on the tungsten converter followed by a 125 mg 226Ra target. The 225Ac is “milked” every few days from the irradiated target in which 225Ra is produced by the photonuclear reaction 226Ra(γ,n)225Ra (15 day half-life). The 225Ac product is pure, i.e. no 227Ac is co-produced. Hence, the 225Ac can be used therapeutically directly or used as a generator of the shorter lived therapeutic alpha emitter 213Bi (46 minute half-life). More than 50 Ci of 213Bi can be extracted via milking the 1 Ci of 225Ac due to the large half-life difference. The 212Pb (10 hour half-life) alpha emitter is the daughter of 224Ra (3.6 day half-life) which is produced at the activity level of ˜4 Ci in the same 5-day irradiation. A total of ˜10 Ci of 212Pb can be milked from the irradiated 226Ra target. That is, the 125 mg 226Ra target irradiated for 5 days as described above can provide >1 Ci of 225Ac or 50 Ci of 213Bi, and simultaneously >10 Ci of 212Pb as the result of successive milkings of the same target using different ion exchange columns. It is to be understood that the above description is intended to be illustrative, and not restrictive. For example, the above-described embodiments (and/or aspects thereof) may be used in combination with each other. In addition, many modifications may be made to adapt a particular situation or material to the teachings of the invention without departing from its scope. While the dimensions and types of materials described herein are intended to define the parameters of the invention, they are by no means limiting, but are instead exemplary embodiments. Many other embodiments will be apparent to those of skill in the art upon reviewing the above description. The scope of the invention should, therefore, be determined with reference to the appended claims, along with the full scope of equivalents to which such claims are entitled. In the appended claims, the terms “including” and “in which” are used as the plain-English equivalents of the terms “comprising” and “wherein.” Moreover, in the following claims, the terms “first,” “second,” and “third,” are used merely as labels, and are not intended to impose numerical requirements on their objects. Further, the limitations of the following claims are not written in means-plus-function format and are not intended to be interpreted based on 35 U.S.C. § 112, sixth paragraph, unless and until such claim limitations expressly use the phrase “means for” followed by a statement of function void of further structure. As will be understood by one skilled in the art, for any and all purposes, particularly in terms of providing a written description, all ranges disclosed herein also encompass any and all possible subranges and combinations of subranges thereof. Any listed range can be easily recognized as sufficiently describing and enabling the same range being broken down into at least equal halves, thirds, quarters, fifths, tenths, etc. As a non-limiting example, each range discussed herein can be readily broken down into a lower third, middle third and upper third, etc. As will also be understood by one skilled in the art all language such as “up to,” “at least,” “greater than,” “less than,” “more than” and the like include the number recited and refer to ranges which can be subsequently broken down into subranges as discussed above. In the same manner, all ratios disclosed herein also include all subratios falling within the broader ratio. One skilled in the art will also readily recognize that where members are grouped together in a common manner, such as in a Markush group, the present invention encompasses not only the entire group listed as a whole, but each member of the group individually and all possible subgroups of the main group. Accordingly, for all purposes, the present invention encompasses not only the main group, but also the main group absent one or more of the group members. The present invention also envisages the explicit exclusion of one or more of any of the group members in the claimed invention.
052415716	claims	1. A zirconium alloy absorber material comprising: erbium in a range of from about 0.05 to 2.0 wt. % selected from the group consisting of a naturally occurring distribution of erbium isotopes, isotopically enriched erbium-167, and a combination thereof; iron in a range from about 0.2 to about 0.5 wt. %; about 50 to 120 ppm silicon; about 1000 to 2200 ppm oxygen; one or more additional alloying metals selected from the group consisting of tin in a range of from a measurable amount up to about 1.4 wt. %, chromium in a range from about 0.07 to about 0.25 wt. %, niobium in a range of from a measurable amount up to about 0.6 wt. %, and vanadium in a range of from a measurable amount up to about 0.5 wt. %; and a balance of zirconium. providing a zirconium alloy having iron in a range from about 0.2 to about 0.5 wt. %; about 50 to 120 ppm silicon; about 1000 to 2000 ppm oxygen; one or more additional alloying metals selected from the group consisting of tin in a range of from a measurable amount up to about 1.4 wt. %, chromium in a range from about 0.07 to about 0.25 wt. %, niobium in a range from a measurable amount up to about 0.6 wt. %, and vanadium in a range of from a measurable amount up to about 0.5 wt. %; and a balance of zirconium; and adding erbium in a range of from about 0.05 to 2.0 wt. % selected from the group consisting of a naturally occurring distribution of erbium isotopes, isotopically enriched erbium-167, and a combination thereof to act as a burnable absorber. 2. The zirconium alloy absorber material of claim 1, wherein said alloy is for use in a nuclear reactor fuel rod cladding tube and wherein said iron is present in an amount of about 0.46 wt. %, said oxygen is present in an amount ranging from about 1600-2200 ppm and said additional alloying metals comprise tin in an amount of about 0.5 wt. % and chromium in an amount of about 0.23 wt. %. 3. The zirconium alloy absorber material of claim 1, wherein said alloy is for use in a nuclear reactor fuel rod cladding tube and wherein said iron is present in an amount of about 0.35 wt. %, said oxygen is present in an amount ranging from about 1000 to about 1200 ppm, and said additional alloying metals comprise tin in a range of from about 0.5 to 1.0 wt. %, chromium in an amount of about 0.25 wt. % and niobium in an amount of about 0.3 wt. %. 4. The zirconium alloy absorber material of claim 1, wherein said alloy is for use in a nuclear reactor fuel rod cladding tube and wherein said iron is present in an amount of about 0.3 wt. %, said oxygen is provided in an amount of about 2200 ppm and said additional alloying metals further comprise vanadium in an amount of about 0.25 wt. %. 5. The zirconium alloy absorber of claim 2, further comprising an outer layer of a more corrosion resistant material in the form of a duplex tubing. 6. The zirconium alloy absorber material of claim 1, wherein said alloy is for use as a structural component in a nuclear reactor and wherein said oxygen is provided in an amount ranging from about 1200 to about 1800 ppm and said additional alloying metals further comprise tin in a range from about 0.5 to about 1.4 wt. %, chromium in a range from about 0.07 to about 0.25 wt. % and niobium in range from about 0.1 to about 0.3 wt. %. 7. The structural component of claim 6, further comprising layers of a more corrosion resistant layer on both the inside and outside surfaces of said structural component. 8. A method of making a zirconium alloy absorber material comprising the steps of:
052251470	claims	1. A method for determining the neutronics parameters of a reactor core comprising the steps of: representing the reactor core as a plurality of nodes having a coarse nodal representation; monitoring selected neutronic parameters of the reactor core; providing time-dependent two group neutron diffusion equations coupled to delayed neutron precursor concentrations that have been subjected to space-time factorization by shape and amplitude functions in response to the plurality of nodes; sensing the monitored parameters; and determining the core neutronics parameters in response to the sensed parameters and the provided two group neutron diffusion equations in constant time steps for sensing the monitored parameters and determining the core neutronics parameters in a real-time environment, the time steps being not less than one quarter second. representing the reactor core as a plurality of nodes; monitoring selected neutronic parameters of the reactor core; providing a solution set of time-dependent two group neutron diffusion equations coupled to delayed neutron precursor concentrations that have been subjected to space-time factorization into shape and amplitude functions for the neutron groups and delayed neutron precursors for the plurality of nodes; sensing the monitored parameters; and solving the shape functions by approximation and solving amplitude functions in response to the sensed parameters in constant time steps, thereby determining the core neutronics parameters. a mathematical model of the reactor core as a plurality of nodes; means for receiving selected neutronic parameters of the reactor core; first processing means for providing a solution set of time-dependent two group neutron diffusion equations coupled to delayed neutron precursor concentrations factorized into shape and amplitude functions for the neutron groups and delayed neutron precursors for the plurality of nodes; and second processing means for solving the shape functions by approximation and for solving amplitude functions in response to the received selected neutronic parameters in constant time steps, thereby determining the core neutrons parameters. 2. The method of claim 1 wherein representing the core further comprises representing the core as a plurality of radial and axial nodes in the range of from 1500 to 4500 total nodes, each radial node having from 8 to 24 axial nodes. 3. A method for determining the neutronics parameters of a reactor core comprising the steps of: 4. The method of claim 3 wherein representing the core further comprises representing the core as a plurality of radial nodes, each radial node having in the range of from 8 to 24 axial nodes. 5. The method of claim 3 wherein solving shape functions and amplitude functions further comprise solving said functions in real time using a first constant time step of not less than 0.25 seconds for solving the amplitude functions and a second constant time step that is a multiple of the first constant time step for solving the shape functions, thereby simulating the full range of operation of the core continuously. 6. The method of claim 3 wherein solving the amplitude functions further comprise applying a dynamic implicit solution method wherein the reactivity is calculated. 7. The method of claim 3 wherein solving the shape functions by approximation further comprises applying the Borresen approximation. 8. The method of claim 3 wherein solving the shape functions further comprises applying a modified Gauss-Seidel approach. 9. Apparatus for determining the neutronics parameters of a reactor core comprising: 10. The apparatus of claim 9 wherein the second processing means comprises a third processing means for solving the shape and amplitude functions for a subset of the plurality of nodes corresponding to a local region of the core, thereby determining the core neutronics parameters for said local core region. 11. The apparatus of claim 9 wherein the plurality of nodes further comprises a plurality of radial nodes wherein each radial node has a plurality of axial nodes selected from between 8 and 24. 12. The apparatus of claim 9 wherein the second processing means operates using a first constant time steps selected on the order of 0.25 seconds for solving the amplitude functions, and a second constant time step that is a multiple of the first constant time step for solving the shape functions, thereby simulating the full range of operation of the core continuously. 13. The apparatus of claim 9 wherein the second processing means further comprises means for solving the amplitude functions by application of a dynamic implicit solution method wherein the reactivity is calculated. 14. The apparatus of claim 9 wherein the second processing means further comprises means for solving the shape functions by application of the Borresen approximation. 15. The apparatus of claim 9 wherein the second processing means further comprises means for solving the shape functions by application of a modified Gauss-Seidel approach. 16. The method of claim 3 wherein the step of providing the solution set further comprises providing the neutron groups with the same amplitude function which is only a function of time. 17. The method of claim 6 wherein the step of providing the solution set further comprise providing the delayed neutron precursors with the same shape function which is a function of time and space. 18. The method of claim 3 wherein the step of providing the solution set further comprises providing the neutron groups with the same amplitude function which is only a function of time. 19. The method of claim 3 wherein the step of providing the solution set further comprises three shape functions and seven amplitude functions. 20. The method of claim 5 wherein the second constant time step is four times the first. 21. The apparatus of claim 9 wherein the neutron groups of the solution set have the same amplitude function which is only a function of time. 22. The apparatus of claim 9 wherein the delayed neutron precursors of the solution set have the same shape function which is a function of time and space. 23. The apparatus of claim 13 wherein the neutron groups of the solution set have the same amplitude function which is only a function of time. 24. The apparatus of claim 9 wherein the solution set further comprises three shape functions and seven amplitude functions.
	abstract	A fastening device for a diaphragm for x-ray radiation has a base plate provided with an opening for the passage of x-ray radiation, the base plate being able to accommodate load forces, and at least one covering device which can be arranged for enclosure of a space on the base plate. A diaphragm can be housed in the space. A computer tomography apparatus embodies such a fastening device for a diaphragm.
048448575	claims	1. Pressurized water reactor with a primary circuit including therein a reactor pressure vessel having a top cover, 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 the top cover of the pressure vessel and having at least one shut-off valve therein, said line connecting the reactor pressure vessel to a part of the auxiliary system wherein a lower pressure prevails than in the reactor pressure vessel. 2. Pressurized water reactor according to claim 1, wherein said line is connected to the suction side of a water-jet pump which is supplied with the water of the auxiliary system as driving fluid. 3. Pressurized water reactor according to claim 1, wherein the auxiliary system is a volume control system of the pressurized water reactor. 4. Pressurized water reactor according to claim 1 wherein the auxiliary system is a boron-addition system of the pressurized water reactor. 5. Pressurized water reactor according to claim 1, wherein a temperature measuring station is located in said line.
056132433	abstract	The specification discloses a process for stabilizing radionuclides extracted during the upgrading of minerals. The process comprises forming a composition of a radionuclide and a component and roasting the composition so that the component forms a crystalline phase having a structure that binds the radionuclides. Suitable components include a compound of a lanthanide and/or phosphorus and zirconia. Zirconia in its cubic form is useful in stabilizing uranium and thorium.
	claims	1. A multi-leaf collimator comprising:a guide frame,a plurality of metal plates arranged in the guide frame in a displaceable fashion,a plurality of piezoelectric motors, each piezoelectric motor including:at least two piezoelectric actuators,an internally toothed driving ring that is excited by a stroke of the at least two piezoelectric actuators to a circulating displacement movement, andan externally toothed shaft attached to the driving ring such that the shaft is rotated by the displacement movement of the driving ring. 2. The multi-leaf collimator according to claim 1, comprising a motor housing that houses a number of the piezo-electric actuators, drive rings and shafts. 3. The multi-leaf collimator according to claim 1, wherein a toothed coupling between each externally toothed shaft and a toothed portion of a corresponding metal plate converts a rotation of the shaft into a linear movement of the metal plate. 4. The multi-leaf collimator according to claim 1, wherein the positioning of the metal plates can be controlled electrically. 5. The multi-leaf collimator according to claim 1, comprising electrical linear transducers for monitoring the position of each metal plate. 6. The multi-leaf collimator according to claim 5, wherein the positioning of the metal plates is electrically controlled, with a signal of the electrical linear transducer of each metal plate being used as a control signal. 7. The multi-leaf collimator according to claim 1, wherein a control electronics system is remote from the electrical motor such that it is arranged in a region with a radiation dose which is lower compared to that of the electrical motor. 8. A method for operating a multi-leaf collimator comprising a guide frame with a plurality of metal plates arranged in a displaceable fashion, and a plurality of piezoelectric motors, each piezoelectric motor including at least two piezoelectric actuators, an internally toothed driving ring that is excited by a stroke of the at least two piezoelectric actuators to a circulator displacement movement, and an externally toothed shaft attached to the driving ring such that the shaft is rotated by the displacement movement of the driving ring, the method comprising:controlling the plurality of piezoelectric motors such that at least one of the metal plates is movable both individually and simultaneously according to individual movement profiles. 9. A method for operating a multi-leaf collimator comprising a guide frame with a plurality of metal plates arranged in a displaceable fashion, and a plurality of piezoelectric motors, each piezoelectric motor including at least two piezoelectric actuators, an internally toothed driving ring that is excited by a stroke of the at least two piezoelectric actuators to a circulating displacement movement, and an externally toothed motor shaft attached to the driving ring such that the motor shaft is rotated by the displacement movement of the driving ring, the method comprising:for each piezoelectric motor, controlling the at least two piezoelectric actuators using a sine/cosine voltage to control the position of the motor shaft, wherein the at least two piezoelectric actuators are arranged at right angles to one another and operate according to a longitudinal effect.
	description	This application is a continuation of International Application No. PCT/CA01/00496, filed on Apr. 17, 2001, which is a continuation-in-part of U.S. application Ser. No. 09/550,923, filed on Apr. 17, 2000, now U.S. Pat. No. 6,504,898. The present invention relates to a method and apparatus for irradiating products to achieve a radiation dose distribution that satisfies specified dose uniformity criteria throughout the product. The treatment of products using radiation is well established as an effective method of treating materials such as medical devices or food stuffs. Radiation processing of products typically involves loading products into totes and introducing a plurality of totes either on a continuous conveyer, or in bulk, into a radiation chamber. Within the chamber the product stacks pass by a radiation source until the desired radiation dosage is received by the product and the totes are removed from the chamber. As a plurality of products, typically within totes, are present in the chamber at a given time, the radiation processing parameters affect all of the product within the chamber at the same time. One common problem in the radiation processing of products is that the effectiveness of radiation processing is sensitive to variations in product density and geometry, and product source geometry. If a radiation chamber is loaded with totes comprising products with a range of densities and geometries,certain products will tend to be over-exposed to the radiation, while others do not achieved the required dose, especially within the central regions of the product. To overcome this problem the radiation chamber is typically loaded with products according to a specified and validated configuration so that the processing of the products satisfies a specified dose uniformity criteria. However, this is not always possible as some product package configurations are not compatible with achieving a good dose uniformity when irradiation is carried out in the conventional manner. Products of a large dimension, and high density suffer from a high dose uniformity ratio (DUR) across the product. A relatively even radiation dose distribution (small DUR) is desirable for all products, but especially so for the treatment of foods, such as red meats and poultry. In treatment of these products, an application of an effective radiation dose to reduce pathogens at the centre of the stack is often limited by associated undesirable sensory or other changes in the periphery of the product stack as a result of the higher radiation dose delivered to material in this region of the product. A similar situation may arise during the radiation sterilization of medical disposable products, a majority of which may be made from plastic materials. In these cases, the maximum permissible radiation dose in a product may be limited by undesirable changes in the characteristics of the plastics, such as increased embrittlement of polypropylene or decoloration and smell development of polyvinyl chloride. In order to adequately and thoroughly treat product stacks of such products with radiation processing, a relatively even radiation dose distribution characterized by a low DUR must be delivered throughout the product stack. Radiation processing of materials and products has most often been accomplished using electron beams, gamma radiation or X-rays. A major drawback to electron beam processing, is that the electron beam is only capable of penetrating relatively shallow depths (i.e. cm) into product, especially high density products such as food stuffs. This limitation reduces the effectiveness of electron beam processing of bulk or palletized materials of high density. Gamma radiation is more effective in penetrating products, especially those of a higher density or larger dimensions, compared with electron beam. Most gamma sources are based on radioactive nuclides such as cobalt-60. Kock and Eisenhower (National Research Council of the National Academy of Sciences Publication #1273; 1965) discuss the merits of different types of radiation processing for the purposes of food treatment. The article suggests that photons are the preferred source for treating large product stacks because of the greater ability of photons to penetrate the product. U.S. Pat. No. 4,845,732 discloses an apparatus and process for producing bremsstrahlung (X-rays) for a variety of industrial applications including irradiation of food or industrial products. An alternate device for the production of X-rays is disclosed in U.S. Pat. No. 5,461,656 which also discloses X-ray irradiation of a range of materials. U.S. Pat. No. 5,838,760 and U.S. Pat. No. 4,484,341 teach a method and apparatus for selectively irradiating materials such as foodstuffs with electrons or X-rays. None of these documents discloses an apparatus or methods to deliver a relatively even radiation dose distribution, especially in large product stacks of high density, so that a low DUR is achieved in treated products. U.S. Pat. No. 4,561,358 discloses an apparatus for conveying articles within a tote (carrier) through an electron beam. The invention teaches of a carrier that is capable of reorienting its position as the carrier approaches the electron beam. An analogous system is disclosed in U.S. Pat. No. 5,396,074 wherein articles are transported past an electron beam on a process conveyor system. The conveyor system provides for re-orientation of the carrier so that a second side (opposite the first side) of the carrier is exposed to the radiation source. The carrier is further defined in U.S. Pat. No. 5,590,602. A similar electron beam irradiation device is disclosed in U.S. Pat. No. 5,994,706. An apparatus to optimize the dosage of electron beam radiation within a product are given in U.S. Pat. No. 4,983,849. The apparatus includes placing cylindrical or plate dose attenuators between the radiation beam and product. The attenuators comprise a moving, perforated metal plate (or cylinder) scatter the radiation beam and reflect non-intersecting electrons thereby increasing dosage uniformity. U.S. Pat. No. 5,554,856 discloses a radiation sterilizing conveyor unit for sterilizing biological products, food stuffs, or decontamination of clinical waste and microbiological products. Products are placed on a disk-shaped transporter and rotated so that the products are exposed to a field of accelerated electrons. A similar apparatus for electron beam sterilization of biological products, foodstuffs, clinical waste and microbiological products is also disclosed in U.S. Pat. No. 5,557,109. Products are placed in a recess or pocket of a manipulator which is slid horizontally into a cavity until the products are aligned with a path of an electron beam housed within the sterilization unit. In the prior art systems described above, there are limitations in the ability to deliver a relatively flat dose distribution (low DUR) throughout a product or product stack since no method is provided to compensate for the different doses received by the exterior and interior portions of the product stack. This therefore results in the outer portions of a product to receive a much higher radiation dose than that received within the product stack. U.S. Pat. No. 4,029,967 and U.S. Pat. No. 4,066,907 disclose an irradiation device for the uniform irradiation of goods by means of electro-magnetic radiation having a quantum energy larger than 5 KeV. Products to be irradiated (including medical articles, feedstuffs, and food) rotate on turntables and are partially shielded from a radiation source by shielding elements. There is no discussion of optimizing the geometry of the radiation beam relative to the product stack, or modifying the spacing of the shielding elements in order to optimize the DUR within a product. As a result, products with different densities are still subject to a wide range in DUR as is the case with other prior art systems. U.S. Pat. No. 5,001,352, also discloses a similar apparatus comprising product stacks that rotate on turntables, positioned around a centrally disposed radiation source, and shielding elements that reduce lateral radiation emitting from the source. A shielding element comprising a plurality of pipes that are fluid filled thereby permitting flexibility in the form of the shielding element is also discussed. However, there is no guidance as to how this or the other shielding elements are to be positioned in order to attenuate the radiation beam relative to the product stack in order to optimize the DUR within the product. Nor is there any discussion of any real-time adjustment of shielding elements to optimize the dose distribution received by a product that accounts for alterations in product densities. A major limitation with the prior art irradiation systems is that it is difficult to obtain a relatively even radiation dose distribution (low DUR) throughout a product or product stack. For example, in systems which irradiate products from only one side, the material irradiated at the periphery of the product and closest to the irradiation source receives a high radiation dose relative to the product located at the center regions of the product stack, and further away from the radiation source resulting in a high DUR. Even with systems that irradiate products from multiple sides, the material irradiated at the periphery of the product typically receives a higher dose of radiation than the material located at the centre of the product since the radiation method is not optimized for the product stacks. Consequently, the product receives an uneven dose of radiation, characterised by a high DUR. Thus, prior art systems are limited in their ability to deliver a relatively flat dose distribution (low DUR) throughout a product or product stack. These limitations are more pronounced in larger products, with higher densities. It is an object of the current invention to overcome drawbacks in the prior art. The above object is met by the combinations of features of the main claims, the sub-claims disclose further advantageous embodiments of the invention. The present invention relates to a method and apparatus for irradiating products to achieve a radiation dose distribution that satisfies specified dose uniformity criteria throughout the product. According to the present invention there is provided a product irradiator comprising: a radiation source, an adjustable collimator, a turntable; and a control system. The radiation source may be selected from the group consisting of gamma, X-ray and electron beam radiation. Preferably, the radiation source is an X-ray radiation source comprising an electron accelerator for producing high energy electrons, a scanning horn for directing the high energy electrons and a converter for converting the high energy electrons into X-rays. The present invention is also directed to the product irradiator as defined above which further comprises a detection system. The detection system measures at least one the following parameters: transmitted radiation, instantaneous angular rotation velocity of the turntable, angular orientation of the turntable, power of the radiation beam, energy of the radiation beam, speed of vertical scan, collimator aperture, width of the radiation beam, position of an auxiliary shield, offset of the radiation beam axis from axis of rotation of the product on the turntable, distance of the turntable from collimator, and distance of collimator from the source. Preferably, the detection system is operatively linked with said control system. The present invention also pertains to a method of radiation processing a product comprising: i) determining length, width, height and density of a product stack comprising the product; ii) determining the width of a collimated radiation beam required to produce a low Dose Uniformity Ratio within the product; iii) adjusting a collimator aperture to obtain the width determined in step ii); and iv) rotating the product stack within the collimated radiation beam for a period of time sufficient to achieve a minimum required radiation dose within the product.This method also pertains to the step of adjusting (step iii), wherein an angular velocity of the turntable may be adjusted. Furthermore, within the step of adjusting, the collimated radiation beam is a collimated X-ray beam produced from high energy electrons generated by an electron accelerator, and power of the high energy electrons may be adjusted. This invention also pertains to the method as defined above wherein during or following the step of rotating, is a step (step v) of detecting X-rays transmitted through the product. Furthermore, during or following the step of detecting (step v), is a step (step vi) of processing information obtained in the detecting step by a control system and altering, if required, of any of the following parameters: collimator aperture, distance between the turntable and collimator, turntable offset, position of auxiliary shield, angular velocity of the turntable, power of the high energy electrons, speed of vertical scan. The present invention also pertains to the use of an apparatus comprising a radiation source for producing radiation energy selected from the group consisting of x-ray, e-beam, and radioisotope, an adjustable collimator capable of attenuating a first portion of the radiation while permitting passage of a second portion of the radiation, the second portion of radiation shaped by the adjustable collimator into a radiation beam, the radiation beam traversing a turntable capable of receiving a product stack, and a control system capable of modulating the adjustable collimator or any one or all irradiation system parameters as the product stack rotates on the turn-table, for delivery of a radiation dose producing a low dose uniformity ratio (DUR) within the product stack The present invention further pertains to a method of irradiating a product stack with a low dose uniformity ratio comprising, rotating a product stack in an X-ray radiation beam of width less than or equal to the diameter of the product stack and modulating the width of the radiation beam relative to the rotating product stack. Modulation of the width of the radiation beam may be effected by adjusting the adjustable collimator, the distance between the product stack and collimator, or the distance between the source and collimator, position of an auxiliary shield, or a combination thereof, as the product stack rotates in the radiation beam. The present invention is directed to a product irradiator comprising: i) an X-ray radiation source essentially consisting of an electron accelerator for producing high energy electrons, a scanning horn for directing the high energy electrons towards a convertor, the converter for converting said high energy electrons into X-rays to produce an X-ray beam, the X-ray beam directed towards a product requiring irradiation; ii) an adjustable collimator for shaping the X-ray beam; iii) a turntable upon which the product is placed; and iv) a control system in operative communication with the electron accelerator, the adjustable collimator and the turntable. This invention also pertains to the product irradiator just defined further comprising a detection system in operative association with the control system. Furthermore, the turntable of the product irradiator may be movable towards or away from the adjustable collimator, or the turntable my be movable laterally, so that an axis of rotation of the product on the turntable is laterally offset from the X-ray beam axis. The product irradiator may also comprising an auxiliary shield. The present invention also pertains to the the product as defined above, wherein the detection system measures at least one the following parameters: transmitted X-ray radiation, instantaneous angular velocity of the turntable, angular orientation of the turntable, power of the high energy electrons, width of high energy electron beam, energy of the X-ray beam, aperture of the adjustable collimator, position of the auxiliary shield, offset of the radiation beam axis from axis of rotation of the turntable, distance of the turntable from collimator, and distance of the collimator from the radiation source. The present invention also pertains to an apparatus for irradiating a product comprising: i) a radiation detection system that measures the amount of radiation absorbed by at least part of the product; ii) a radiation source; iii) a collimator, and iv) a turntable.wherein each of the source, collimator and turntable have at least one parameter that is capable of being adjusted automatically based upon a measurement made by the detection system to achieve a low Dose Uniformity Ratio in a product during irradiation. The present invention embraces a medium storing instructions adapted to be executed by a processor to modulate either: i) the width of a collimator while a product is being rotated by a turntable, and irradiated by a radiation beam, wherein the radiation beam is collimated by the collimator; ii) the intensity of a radiation beam while a product is being rotated by a turntable, and irradiated by the radiation beam; iii) the rate of rotation of a turntable table, while a product is being irradiated by the radiation beam; and iv) optionally, modifying the vertical scan speed. The present invention also provides for a system for irradiating a product comprising; i) means for producing a radiation beam; ii) means for measuring the amount of radiation absorbed by at least part of the product; iii) means for adjustably setting the width of the radiation beam that irradiates the product; iv) means for rotating the product; v) means for modulating the rate of rotation of the product, modulating the adjustable width of the radiation beam during irradiation based upon the measured amount of radiation absorbed by at least a part of the product.Furthermore, the present invention relates to the system described above further comprising means for modulating intensity of the radiation beam based upon the measured amount of radiation absorbed by at least part of the product. This summary of the invention does not necessarily describe all necessary features of the invention but that the invention may also reside in a sub-combination of the described features. The present invention relates to a method and apparatus for irradiating products to achieve a radiation dose distribution that satisfies specified dose uniformity criteria throughout the product. The following description is of a preferred embodiment by way of example only and without limitation to the combination of features necessary for carrying the invention into effect. By “radiation processing” it is meant the exposure of a product, or a product stack (60) to a radiation beam (40; FIG. 4; or 45; FIG. 5) or a collimated radiation beam (50; FIGS. 4 to 6). The product must be within the radiation chamber (80), and the radiation source must be placed into position and unshielded as required to irradiate the product, for example as in the case of but not limited to a radioactive source (100; for example the radioactive source that is raised from a storage pool), or the radiation source must be in an active state, for example when using an electron-beam (15), or X-rays derived from an electron beam (e.g. 45; FIG. 5) in order to irradiate the product or product stack (60). It is to be understood that any product may be processed according to the present invention, for example, but not limited to, food products, medical or laboratory supplies, powdered goods, waste, for example biological wastes. By the term “dose uniformity ratio” or “DUR” it is meant the ratio of the maximum radiation dose to the minimum radiation dose, typically measured in Grays (Gy) received within a product or product stack, and is expressed as follows:DUR1=Dosemax/Dosemin Dosemax (also referred to as Dmax) is the maximum radiation dose received at some location within the product or product stack in a given treatment, and Dosemin is the minimum radiation (also referred to as Dmin) dose received at some location within the same product or product stack in a given treatment. A DUR of 2 indicates that the highest radiation dose received in a volume element located somewhere within the product stack is twice the lowest radiation dose delivered in a volume element located at a different position within the same product or product stack. A DUR of about 1 indicates that a uniform dose distribution has been delivered throughout the product material. A “high DUR” is defined to mean a DUR greater than about 2. A “low DUR” is defined to mean a DUR of about 1 to less than about 2. These are arbitrary categories. Conventional irradiation systems are characterized as producing a high DUR of above 2 for low density products, and above 3 for products with densities greater than or equal to 0.8 g./cm3. By the term “accelerator” (20; FIG. 5) it is meant an apparatus or a source capable of providing high energy electrons preferably with energy and power measured in millions of electron volts (MeV) and in kilowatts (kW) respectively. The accelerator also includes associated auxiliary equipment, such as a RF generator, Klystron, power modulation apparatus, power supply, cooling system, and any other components as would be known to one skilled in the art to generate an electron beam. By the term “scanning horn” it is meant any device designed to scan a beam of high energy electrons over a specified angular range. The dimensions may include a horizontal or a vertical plane of electrons. The scanning horn may comprise a magnet, for example, but not limited to a “bowtie” magnet, to produce a parallel beam of electrons emitting from the horn. Also, the “scanning horn” may be an integral part of the accelerator or it may be a separate part of the accelerator. By the term “converter” (30; FIG. 5) it is meant a device or object designed to convert high energy electrons (10, 15) into X-rays (45; FIG. 5). By the term “collimator” or “adjustable collimator” (110) it is meant a device that shapes a radiation beam (40, 45) into a desired geometry (50). Typically the shape of the radiation beam is adjusted in its width, however, other geometries may also be adjusted, for example, but not to be considered limiting, its height or both its height and width, as required. It is also contemplated that non-rectangular cross-sections of the beam are also possible. The collimator defines an aperture through which radiation passes. The collimator may have a shallow profile as depicted in FIG. 3(a), or may have an elongated profile as depicted in FIG. 3(b). An elongated collimator, such as that shown in FIG. 3(b) helps focus the radiation beam by altering the penumbra. Adjustments to the aperture of the collimator shape the radiation beam into the desired geometry and dimension required to produce a DUR approaching 1 for a product stack with particular characteristics (such as geometry and density). By the term “adjustable collimator” it is meant a collimator with an adjustable aperture that shapes the radiation beam into any desired geometry, for example, but not limited to adjusting the height, width, offset of the beam axis from the axis of rotation of the turntable, or a combination thereof, before or during radiation processing of a product or product stack. For example, an adjustable collimator may comprise a two or more radiation opaque shielding elements (for example, 115), that move horizontally thereby increasing or decreasing the aperture of the collimator as required. Shielding elements other than that shown in FIGS. 4 to 6 may also be used that adjust the aperture of the collimator. For example, which is not to be considered limiting, the shielding elements may comprise a plurality of overlapping plates each being radiation opaque, or partially radiation opaque, and capable of moving independently of each other. The overlapping plates may be moved as required to adjust the opening of aperture 170 (see Examples 2 and 3 for results relating to optimizing DUR by adjusting aperture width of collimator). The shielding elements may also comprise, which again is not to be considered as limiting, a plurality of pipes (e.g. U.S. Pat. No. 5,001,352; which is incorporated herein by reference) each of which may be independently filled, or emptied, with a radiation opaque substance. The filling or emptying, of the pipes adjusts the effective width of the collimator aperture as required. By “auxiliary shield” it is meant a device that partially blocks the radiation beam and is placed within the radiation beam, between the converter and product stack (see 300, FIGS. 3(d) and 13(a), Example 4). The auxiliary shield helps to further shape the radiation beam, regulate penumbra, and reduce the dose at the center of the radiation beam within the product stack The auxiliary shield may be movable along the axis of the radiation beam so that it may be variably positioned in the path of the radiation beam, between the converter and product stack. Auxiliary shields that are appropriately shaped, and that may span the entire collimator aperture are also effective in reducing DUR, for example, but not limited to those shown in FIG. 13(a). By the term “detection system” (130) it is meant any device capable of detecting parameters of the product stack before, and during radiation processing. The detection system may comprise one or more detectors, generally indicated as 180 in FIG. 6, that measure a range of parameters, for example but not limited to, radiation not absorbed by the product. If measuring transmitted radiation, such detectors are placed behind the product to measure the amount of radiation transmitted through the product stack. However, detectors may also be placed in different locations around the product, or elsewhere so that other non-absorbed radiation is monitored. Other detectors may also be used to determine parameters before, or during radiation processing, including but not limited to those that measure the position of rotation of the turntable (angular orientation), instantaneous angular velocity of the turn table, collimator aperture, product density, product weight, product stack dimensions, energy and power of the electron beam, and other parameters associated with the conveying system or geometry of the system arrangement. A control system, generally indicated as 120 in FIG. 7, is used to receive the information obtained by the detector system (130) to either maintain the current system settings, or adjust one or more components of the irradiation system of the present invention as required (see FIG. 6). These adjustments may take place before, or during radiation processing of a product. Components that are monitored by the control system (120), and that may be adjusted in response to information gathered by the detector system (130) include, but are not limited to, the size of aperture (170, i.e. the beam geometry), power of the radiation beam (45), energy of the radiation beam (15), speed of rotation of the turntable (70), angular position (orientation) of turntable (230), instantaneous angular velocity of the turntable, distance of the collimator from the source (‘L’, FIG. 3(a); 220, FIG. 7), distance of the turntable from the collimator (‘S’, FIG. 3(a); 250, FIG. 7), and conveying system (150). In this manner, the control system (120) uses parameters derived from characteristics obtained from the detector system (130) in order to optimize the radiation dose distribution delivered to the product stack (60). The control system includes, in addition to the detection system (130), hardware and software components (120) required to process the information obtained by the detector system, and the interfacing (200, 210) between the computer system (120) and the detector system (interface 200), and the elements of the radiation system (interface 210). Theory for Optimizing DUR within a Product Stack FIG. 1, illustrates the radiation dose profiles within a product that has been exposed to irradiation from either one or two sides which are common within the art. for example, irradiation processes involving one side are disclosed in U.S. Pat. Nos. 4,484,341; 4,561,358; 5,554,856; or 5,557,109. Similarly, two-sided irradiation of a product is described in, for example, U.S. Pat. Nos. 3,564,241; 4,151,419; 4,481,652; 4,852,138; or 5,400,382. Shown in FIGS. 1(a) and (c) are two dimensional representations of the irradiation of a product stack from a single side with a uniform radiation beam. The radiation dose delivered through the depth of the product along line X–X′ of FIGS. 1(a) and (c) is represented in FIGS. 1(b) and (d), respectively. The dose response curve decreases with distance from the product surface nearest the source to a minimum level (Dmin) at the opposite side of the product, With one sided radiation processing the DUR (Dmax/Dmin) is much greater than 1. ‘D’ represents the minimum radiation dose required within the product for a desired specific effect, for example but not limited to, sterilization. A portion Of the product has not reached the minimum required dose in FIG. 1(b) therefore a longer irradiation period is required for all of the product to reach at least the minimum required dose (D). This results in over exposure of the product on the side facing the radiation source and this is undesirable for the processing of many products that are modified as a result of exposure to excessively high doses of radiation. Similar modeling for two sided irradiation of a product is presented in FIGS. 1(e) and (f). Under this radiation processing condition two sides of the product receive a high radiation dose, relative to the middle of the product. Two sided irradiation still results in a relatively high DUR in the product, but the difference between Dmax and Dmin is reduced, and the DUR is improved when compared to one-sided irradiation. FIG. 2(a), illustrates a two dimensional view of the irradiation of a product rotating about its axis in a uniform radiation field where the width of the radiation beam is greater than or equal to the diameter of the product. The product for simplicity is depicted as having a circular cross section, however, rectangular products, or irregularly shaped products may also be rotated to produce similar results as described below. Shown in FIG. 2(b) is the corresponding radiation dose profile received by the product shown along line X–X′. Under these conditions, the radiation dose distribution delivered in the product along X–X′ approximates the radiation dose distribution delivered to the product in two-sided radiation (also along X–X′; FIG. 1(e)) resulting in relatively high DUR. If a rotated product is irradiated using a radiation beam that is much narrower than the diameter (or maximum width) of the product, and which passes through the centre of the product as shown in FIG. 2(c), then the radiation dose distribution curve along X–X′ is relatively low at the periphery of the product and much greater at the centre of the product (see FIG. 2(d)). In such a case, the centre of the product is always within the radiation beam, whereas volume elements such as those defined by points R1 and R2 (FIG. 2(c)) only spend a portion of time in the radiation beam. This fractional exposure time is a function of ‘r’ (FIG. 3(a)) and beam width (‘A’, FIG. 3(a)). The beam width can be controlled in order to control fractional exposure time and hence dose within the product. The fractional exposure time may also be controlled by offsetting the beam from the central axis of rotation of the product (see FIG. 3(c)). Both radiation dose distribution curves (FIGS. 2(b) and (d)) exhibit large differences between Dmax and Dmin and the DUR of these products is still much greater than 1. However, by using a radiation beam wider than the product, or a radiation beam much narrower than the product, the dose distribution profile within the product can be inverted. Therefore, an optimal radiation beam dimension relative to a rotating product such as that shown in FIG. 2(e) can be determined, which is capable of irradiating a rotating product and producing a substantially uniform dose throughout the product with a DUR approaching 1 (FIG. 2(f)). It is also to be understood that by varying the diameter of the incident radiation beam, for example, by altering the width of the scanning pattern, that the penumbra (390) of the beam may be altered. Typically by increasing the beam width, the penumbra also increases (see FIG. 3(a)). The primary beam intensity and penumbra may also be modulated by placing an auxiliary shield (300) between the converter and product (e.g. FIG. 3(d)). Auxiliary shields may block X-ray transmission, or be partially translucent with respect to the transmission of X-rays, for example shields may comprise, but are not limited to, Al or Ta (see Example 4). Furthermore, the auxiliary shield may comprise a variety of shapes, for example, but not limited to shields having a circular, rectangular or triangular cross section, and may span a variety of widths of the aperture (examples of shapes of auxiliary shields are provided in FIG. 13(a)). By inserting an auxiliary shield in the path of the X-ray beam, the central region with a product receives a lower dose, lowering the DUR. Without wishing to be bound by theory, a Ta auxiliary shield may filter the X-ray beam and only permit X-rays of high energy to enter the product (i.e. harden the X-ray spectrum). Another method for altering the dose received within the product is to offset the position of the radiation beam axis with respect to the product axis of rotation (FIG. 3(c)). In this arrangement, a portion of the product is always out of the radiation beam as the product rotates, while the central region of the product receives a continual, or optionally reduced, radiation dose. An example of offset of about 7 cm from the center of rotation, which is not to be considered limiting in any manner, is provided in Example 5. Using an offset, a DUR of 1.4 to about 1.2 may be obtained. The optimal beam dimension must also account for other factors involved during radiation processing, for example but not limited to, product density, the size of aperture (170, i.e. the beam geometry), power of the radiation beam (45), energy of the radiation beam, vertical scan speed as a function of vertical position (instantaneous vertical scan speed), speed of rotation of the turntable (70), angular position (orientation) of turntable (230), instantaneous angular velocity of the turntable, distance of the collimator from the source (‘L’; 220), and distance of the turntable from the collimator (‘S’; 250; also see FIG. 7). Irradiation Parameters Affecting DURs in Products As indicated above, the ratio of the radiation beam width, as determined by the aperture (A), to the width (or diameter) of the product (r) is an important parameter for obtaining a low DUR within a product. As shown in FIG. 2(d), for products of uniform density, the smaller the ratio of A/r, the higher the accumulated dose is at the centre of the stack relative to that at the periphery. Conversely, the larger the ratio of A/r, the accumulated dose is greater at the stack periphery (FIG. 2(b)). In the case of a cylindrical product, the optimum ratio of A/r, producing the lowest DUR within the product, can be constant (FIG. 2(f)). However, in the case of a rectangular product, such as is found in most pallet loads, the effective principal dimension is a function of its angular position () with respect to the beam, since the width of the product changes as the product rotates. Therefore, to maintain an optimal DUR within the product, the ratio of A/r is adjusted as required. For example the Air ratio may be determined for a product of known size and density, so that ‘A’ is set for an average ‘r’. This determination may be made based on knowledge of the contents, density and geometry of the product (or tote), and this data entered into the system prior to radiation processing, or it may be determined from a diagnostic scan (see below; e.g. FIG. 6) of a product prior to radiation processing. It is also contemplated that the A/r ratio may be modulated dynamically as a rectangular product rotates in the radiation beam. The A/r ration may be adjusted by either modifying the aperture (170) of the collimator (110), by adjusting the diameter of the beam (i.e. adjusting beam width, and modulating penumbra), by moving shielding elements 115 appropriately, by placing an auxiliary shield (300) between the converter and product, by moving turntable 70 as required into and away from the source, by adjusting the aperture, offset, and modifying the turntable distance from the source, or by adjusting the distance, ‘L’, between the collimator (110) and source (100). The geometry of the radiation beam (40, 45) produced from a source, for example, but not limited, to a γ-radiation (40) emitted by a radioactive source (e.g. 100; for example but not limited to Co-60), or accelerating high energy electrons (10, 15) interacting with a suitable converter (30) to produce X-rays (45), is determined by the relationship between the following parameters: a) the width of the radiation beam, either γ or X-ray (50; FIG. 3); b) the distance (L) between the source (100) or converter (30) and the collimator (110); c) the distance (S) between the collimator (110) and the product (60) center of rotation, d) the size of the aperture (A) in the collimator (110), and e) the position of an auxiliary shield (300).These parameters determine divergence of the beam and the associated penumbra. Optimisation of these parameters relative to the size and density of a product reduces the DUR within the product.Dynamically Adjusting ‘A/r’ and Associated Parameters During Processing An initial adjustment of the ratio of beam width to the product width (A/r) for a product of a certain density is typically sufficient for a range of product densities and product configurations to obtain a sufficiently low DUR. However, in the case of irregular, or irregular rectangular product shapes, or product containing products with differing densities, modulation of the A/r ratio may be required to obtain a low dose uniformity within a product. Other parameters may also be adjusted to optimize dose uniformity within the product. These parameters may include adjustment of the speed of rotation of the product, modifying the beam power, thereby modulating the rate of energy deposition within the product, or both. Modulation of beam power may be accomplished by any manner known in the art including but not limited to adjusting the beam power of the accelerator, or if desired, when using a radioactive isotope as a source, attenuating the radiation beam by reversibly placing partially radiation opaque shielding between the source and product. Minor adjustments to the intensity of the radiation beam may also include modulating the distance between the product and source. Design of the converter (30) also may be used to adjust the effective energy level of an X-ray beam. As the thickness of the converter increases, lower energy X-rays attenuate within the converter, and only X-rays with high energy exit the converter. Therefore by varying the thickness of the converter the energy level of all, or of a portion of, the X-ray beam may be modified. For example, in the case where the electrons emitting from the scanning horn are not parallel, it may be desired that the upper and lower regions of the X-ray beam be of higher average energy since the beam travels through a greater depth within the product, compared to the beam intercepting the mid-region of the product (however, it is to be understood that parallel electrons may be produced from a scanning horn using one or more magnets positioned at the end of the scanning horn to produce a parallel beam of electrons). Furthermore, these regions of the product experience less radiation backscatter due to the abrupt change in density at the top and bottom of the product. Therefore, a converter with a non-uniform thickness, wherein the thickness increases in its upper and lower portions, may be used to ensure higher energy X-rays are produced in the upper and lower regions from the converter. Modifications to converter thickness typically can not be performed in real time. However, different converters may be selected with different thickness profiles that correspond with different densities or sizes of products to be processed. Furthermore, the power of the beam may also be modulated as a function of vertical position within the product so that a higher power is provided at the upper and lower ends of the product. Additionally, the scan speed of the electron beam can be varied as a function of position of the beam relative to the converter, product, or both the converter and product. If a constant rate of scan of the electron beam is maintained, then due to the scatter of the X-rays produced from the converter, higher levels of radiation are delivered within the central area of the product, and decreasing amounts of radiation are delivered at the ends of the product. An example of the variation is the dose delivery within the vertical dimension of a product can be seen as a solid line in FIG. 3(e). In this example, the bottom and top regions of the product receive about 50% of the radiation when compared to the central region of the product. This variation may be reduced in a variety of ways, examples of which include and are not limited to, modulating the speed of the beam in the “Z” (vertical) direction relative to the product (which may be stationary in the vertical direction), or moving the product vertically relative to the beam, which may be stationary, increasing the relative duration of irradiation at the upper and lower regions of the product, modifying the instantaneous vertical scan speed, using a smaller scan horn thereby reducing the scatter of the X-ray beam, or using a smaller aperture height, again reducing scatter of the X-ray beam. This latter alternative may be obtained by increasing the rate of vertical scan when the electron beam is delivering energy within the mid-vertical region of the product, and reducing the rate of scan towards each of the extremities of the vertical scan (at both the top and bottom of the product). In this manner, the amount of radiation received at the top and bottom regions of the product is increased, while the central dose is decreased somewhat (dashed line, FIG. 3(e)). Other methods may be employed to increase the effective dose received at the ends (upper and lower) of the product. Since the upper and lower regions of the product experience less radiation backscatter, the density discontinuity at these regions may be reduced or eliminated by placing reusable end-caps of substantial density onto the turntable and top of the product as required, thereby increasing back-scatter at these regions. Referring now to FIG. 4, which illustrates an embodiment of the present invention, a radiation source (100) provides an initial radiation beam (40) of an intensity and energy useful for radiation processing of a product. The radiation source may be a radioactive isotope, electron beam, or X-ray beam source. Preferably, the source is an X-ray source produced from an electron beam (see FIGS. 5 and 6). The radiation beam passes through the aperture (generally indicated as 170) of an adjustable collimator (110) to shape the initial radiation beam (40) produced by the radiation source (100) into a collimated radiation beam (50). The aperture of the collimator can be adjusted to produce a collimated radiation beam of optimal geometry for radiation processing a product (60) of known size and density. The distance between the product and the source, collimator, or both source and collimator (e.g. L and S; FIG. 3) may also be adjusted as required to optimize the A/r ratio, and hence the DUR, for a given product. The product (60) rotates on turn table (70) in the path of the collimated radiation beam (50). The product rotates at least once during the time interval of exposure to the radiation source. Preferably, the product rotates more than once during the exposure interval to smooth any variation of dose within the product arising from powering up or down of the accelerator. Detectors (180), and turn-table (70) are connected to the control system (120) so that the size of the aperture (170) of the adjustable collimator (110), the power (intensity) of the initial radiation beam (40), the speed of rotation of turntable (70), the distance of the turntable from the source (L+S), collimator (S), or a combination thereof, may be determined and adjusted, as required, either before or during radiation exposure of the product (60). The embodiment described may also be used to irradiate products (60) of known. dimensions and densities and achieve a relatively low DUR within the product. As one skilled in the art would appreciate, the radiation dose being delivered to the product may be varied as required to account for changes in the distance of the product to the source, width of the rotating product, and density of product. For example, but not to be considered limiting, control system (120) may comprise a timer which dynamically regulates the aperture (170) of adjustable collimator (110) to produce a collimated radiation beam of controlled width (A), to account for changes in the width (r) of rotating product (69). The beam power of radiation source (100) may also be modulated as a function of the rotation of turn-tables (70; as detected by angular position detector 230). In such a case, for example, but which is not to be considered limiting, a rectangular product of known dimension may be aligned on turn-table (70) in a particular orientation (detected by 230) such that as turn-table (70) rotates through positions which bring the corners of the product closer to radiation source (100) the radiation beam may be modified. Such modification may include dynamically adjusting the collimator (110) to modulate the dimension (e.g. A) of the collimated radiation beam (50), adjusting the width of the beam diameter, for example by adjusting the width of the scanning pattern, adjusting the distance between the product and source, or collimator, thereby modifying the relative beam dimension (A) and energy level with respect to the product, or placing or positioning an auxiliary shield (300) between the converter and product in order to adjust penumbra, and to shield and reduce the central dose of the radiation beam within the product. The control system may also regulate the energy and power of the initial radiation beam. Alternatively, control system (120) may regulate the rotation velocity of the turn-table as it rotates thereby allowing the corners of the product to be irradiated for a period of time that is different than that of the rest of the product. It is also contemplated that the control system may dynamically regulate any one, or all, of the parameters described above. Referring now to FIG. 5, which illustrates another embodiment of the invention, wherein radiation source (100) is a source of X-rays produced from converter (30). Electrons (10) from an accelerator (20) interact with a converter (30) to generate X-rays (45). The X-ray beam (45) is shaped by aperture (170) of adjustable collimator (110) into a collimated X-ray beam (50) of optimal geometry for irradiation of the product (60) which rests on turn-table (70). Again, control system (120) monitors and, optionally, controls several components of the apparatus, including the rotation of turn-table (70), aperture of the collimator (110), power of the electron beam produced by accelerator (20), distance between turntable and the collimator (L), or a combination thereof. During radiation processing, product (60) rotates about its vertical axis and intercepts a vertical collimated radiation beam (50). The product rotates at least once during the time exposed to radiation. In most, but not all instances, the width (A; FIG. 3) of the collimated beam is relatively narrow compared to the width of the product (r). Since the vertical plane of the collimated beam (50) is aimed at the centre of the rotating product (60), the periphery of the product is intermittently exposed to the radiation beam. This arrangement compensates for the relatively slow dose build-up at the centre of the product due to attenuation of X-rays by the materials of the product and produces a low DUR. With increased product density, for example but not limited to food such as meat, a narrower collimated beam width will be required in order to obtain a low DUR. Conversely, if a product is of a lower density (for example, medical supplies or waste) the beam width may be increased, or the radiation beam offset from the axis of rotation of the product, since the central portion of the product will receive its minimum dose more readily than that of a product of higher density. In the embodiment shown in FIG. 5, the control system (120) is capable of modulating any or all of the irradiation parameters as outlined above. In certain cases however, such as irradiation of cylindrical products of uniform and relatively low densities, for example sterilization medical products, or it may be advantageous to irradiate the product with a radiation beam having a width approaching or approximately equal to the width of the product. The adjustable collimator of the proposed invention effectively allows this to be accomplished. By controlling the processing parameters this basic principle permits a relatively uniform radiation dose distribution and thus a low DUR to be delivered throughout the product for a large range of product size, shape and densities. The converter (30) may comprise any substance which is capable of generating X-rays following collision with high energy electrons as would be known to one of skill in the art. The converter is comprised of, but not limited to, stainless steel, or high atomic number metals such as, but not limited to, tungsten, tantalum, gold or mercury. The interaction of high energy electrons with converter (30), produces X-rays and heat. Due to the large amount of heat generated in the converter material during bombardment by electrons, the converter needs to be cooled with any suitable cooling system capable of dissipating heat. For example, but not wishing to be limiting, the cooling system may comprise one or more channels providing for circulation of a suitable heat-dissipating liquid, for example water, however, other liquids or cooling systems may be employed as would be known within the art. The use of water or other coolants may attenuate X-rays, and therefore the cooling system needs to be taken into account when determining the energy level of the X-ray beam. As indicated above, attenuation of X-rays within the converter affects the energy spectrum of X-rays escaping from the converter. For example, which is not to be considered limiting, a tantalum converter of about 1 to about 5 mm thickness, with a cooling channel covering the downstream side of the converter, may be used to generate the bremsstrahlung energy spectrum for product irradiation as described herein. The cooling channel may comprise, but is not limited to two layers of aluminum, defining a channel for coolant flow. FIG. 6 illustrates another embodiment of the present invention, where electrons (10) from an accelerator (20) interact with a converter (30) to generate X-rays (45). The X-rays (45) are shaped by aperture (170) of adjustable collimator (110) into an X-ray beam (50) of optimal geometry for irradiation of a product. Transmitted X-Rays (140) passing through product (60) are detected by one or more detector units (180). Detection system (130) is connected with detector units (180) and other detectors that obtain data from other components of the apparatus including turntable rotation velocity (70) and angular position (230), distance between turntable and collimator (S; 250, FIG. 7), accelerator power (20), collimator aperture width (170), conveyor position, via interface 200 and 210. The detection system (130) also interfaces with control system (120; FIG. 7) which also comprises a computer (120) capable of processing the incoming data obtained from the detectors, and sending out instructions to each of the identified components to modify their configuration as required. Detector units (180) may comprise one or more radiation detectors for example, but not limited to, ion chambers placed on the opposite side of the product (60) with respect to the incident radiation beam (50). As the product turns through the radiation beam (50) the detector units (180) register the transmitted radiation dose rate. The difference between incident and exiting radiation dose, and its variation along the stack height is related to the energy absorbing characteristics of the product as a function of several parameters for example, energy of the radiation beam, distance between the turntable (product) and the collimator (S), as a function of the product's angular position. The difference can thus be directly related to the density and geometry of the product. This information may also be used for obtaining a diagnostic scan (see below) of the product. An example of detector arrays that may be used in the system just described is disclosed in WO 01/14911 (which is incorporated herein by reference). A schematic representation of the control system (120) as described above is show in FIG. 7. The control system (120) comprises a computer capable of receiving input data, for example the required minimum radiation dose for a product (190), and data from components of the detection system (180) comprising the accelerator (240), turntable speed of rotation (70), angular position (230), distance to collimator (220), collimator aperture (170), wedge in and out location (310), beam axis offset (280), beam diameter and amplitude (260), and conveyors (150). The control system also establishes settings for, and sends the appropriate instruction to, each of these parameters to optimize properties of the radiation beam relative to the product and produce a low DUR. Those of skill in the art will understand that variations of the control system may be possible without departing from the spirit of the current invention. The embodiment outlined in FIG. 6 permits real-time monitoring of radiation processing of a product, and for real time adjustment between radiation processing of products that differ in size, density or both size and density, so that an optimal radiation dose is delivered to each product to produce a low DUR. Adjustments to the parameters of the apparatus described herein may be made based on information obtained from a diagnostic scan. An optimized radiation exposure may be determined by calculating the difference between the transmitted radiation detected by detector units (180) and the incident radiation at the surface of the product closest to the radiation source (this value can be calculated or determined via appropriately placed detectors), as a function of the rotation of the product. In this way, the radiation dose of any product may be “fine-tuned” to deliver a requisite radiation dose to achieve a low DUR within a product. The inclusion of a radiation detection system (130) also permits obtaining a diagnostic scan of the product (60) to determine the irradiation parameters required to deliver a relatively even radiation dose distribution (low DUR) in a product. The diagnostic scan characterises the product (60) in terms of its geometry and apparent density before any significant radiation dose is accumulated in the product. As suggested in previous embodiments described herein, the diagnostic scan is not required for products of uniform density and stack geometry. The diagnostic scan may be carried out during the first turn of the product (60), or the diagnostic scan may be performed during multiple rotations of the product. The diagnostic scan may comprise irradiating the product with a low power beam so that a low dose is received within the product, for example, but not limited to from about 1 to about 50% of the maximum radiation dose to be received by the product. However, it is to be understood that higher doses may also be used for the diagnostic scan if required. The difference in the amount of radiation sent to the product, and that transmitted through the product (as detected by detectors 130) gives an indication of the density and uniformity of the product. The information determined as a result of the diagnostic scan may be used to set the operational parameters as described herein for product irradiation. Those skilled in the art would understand that in order to irradiate a product to obtain a low DUR, the radiation beam must be capable of penetrating at least to the midpoint of a product. Similarly, if the detection system of the current invention is employed to automatically set the parameters for radiation processing of the product, then the radiation must be capable of penetrating the product. The control system (120) of the present embodiment is designed to simultaneously adjust any one or all the processing parameters of the apparatus as described herein, for example but not wishing to be limiting, the total radiation exposure time, the ratio of the radiation beam width to the principal horizontal dimension of the product, in relation to the angular position (φ) of the X-ray beam (ratio of A(φ)/r(φ)), the power of the radiation beam, the rotational velocity of the turn-table, and the distance between the product and collimator. The control system may adjust the processing parameters based on the total radiation dose required within the product as input by an operator, or the radiation dose may be automatically set at a predetermined value. For example, but not wishing to be limiting, if it is known that a certain base radiation dose is required for a given product, for example the treatment of a food product, then this dose may be preset, and the operating conditions monitored to achieve a low DUR for this dose. However, if two products are of different dimensions or different densities then dissimilar irradiation parameters may be required to deliver the predetermined total radiation dose with an optimal DUR to each stack. As shown in FIG. 8(a), the apparatus of the present invention may be placed within a conveyor system to provide for the loading and unloading of products (60) onto turntable 70. A conveyor (150) delivers and takes away products, for example but not limited to, palletized products or totes, to and from the turntable (70). In the embodiment shown, the collimated radiation beam is produced from a converter (30) that is being bombarded with electrons produced by accelerator 20, and travelling through a scanning horn (25). However, it is to be understood that the source may also be a radioactive isotope as previously described. Not show in FIG. 8(a) are components of the detection or control systems. An outline of a series of process involved in irradiating a product using the methods as described herein is provided, but not limited to, the sequence in FIG. 8(b). Typically, a product (60; FIG. 8(a)) is received and the quality of the product, or product stack determined by any suitable means, for example, by visual inspection. If the product stack is of poor quality the stack is repaired or re-stacked. The product is transported to, and positioned on the turntable, where the product is characterized using one or more characteristics of the product, for example, but not limited to product weight, product dimension, a diagnostic scan wherein the product is characterized in terms of one or more properties, for example, but not limited to, its geometry and apparent density so that the mass distribution through the product may be determined, or a combination thereof. From this product characterization, and the desired dose to be delivered to the product, and the processing protocol (see FIG. 8(c)) is determined to minimize the DUR. The parameters considered in selecting control functions (to create the processing protocol) that determine the dose to be given to a product are shown in FIG. 8(c). The processing protocol is dependent upon product characteristics, and the aperture of the collimator, speed of rotation of the turntable (instantaneous rotational velocity), power of the radiation beam, duration of treatment time, or other variables as described herein (see FIGS. 7 and 8(c)). These parameters may be stored in any suitable manner, for example, within the memory of the control system or on a disc or other suitable medium as desired. Once these parameters are established and the components of the product irradiator set, the product is treated with radiation for a period of time. Preferably, the treatment takes place in the same location as the diagnostic scan, however, the diagnostic scan and creation of the processing protocol (selection of control functions, and storage of appropriate instructions) outlined in FIG. 8(c) may take place at a first location, and the product moved to a second location for irradiation using the processing protocol created as outlined in FIG. 8(c). Therefore, the present invention also provides a medium storing instructions adapted to be executed by a processor to modulate parameters involved during product irradiation. These parameters may include, but are not limited to, one or more of: the width of a collimator, modulation of the intensity of a radiation beam, modulation of the scan speed, modulation of the rate of product rotation, and the exposure time. The duration of treatment may be predetermined and derived from the step of product characterization, for example using a diagnostic scan, or the radiation may be monitored in real-time during treatment using detector units (180, FIG. 6). When the desired radiation dose is obtained, and the product treated, the product is then transported from the turntable to an unload-area. A report recording the processing parameters of the treatment may be generated by the control system (120) as required. Products to be processed using the apparatus and method of the present invention may comprise foodstuffs, medical articles, medical waste or any other product in which radiation treatment may promote a beneficial result. The product may comprise materials in any density range that can be penetrated by a radiation beam. Preferably products have a density from about 0.1 to about 1.0 g/cm3. More preferably, the range is from about 0.2 to about 0.8 g/cm3. Also, the product may comprise but is not necessarily limited to a standard transportation pallet, normally having dimensions 42×48×60 inches. However any other sized or shaped product, or product may also be used. The present invention may use any suitable radiation source, preferably a source that produces X-rays. The electron beam may be produced using an RF (radio frequency) accelerator, for example a “Rhodotron” (Ion Beam Applications (IBA) of Belgium), “Impela” (Atomic Energy Of Canada), or a DC accelerator, for example, “Dynamitron” (Radiation Dynamics), also the radiation source may produce X-rays, for example which is not to be considered limiting, through the ignition of an electron cyclotron resonance plasma inside a dielectric spherical vacuum chamber filled with a heavy weight, non-reactive gas or gas mixture at low pressure, in which conventional microwave energy is used to ignite the plasma and create a hot electron ring, the electrons of which bombard the heavy gas and dielectric material to create X-ray emission (U.S. Pat. No. 5,461,656). Alternatively, the radiation source may comprise a gas heated by microwave energy to form a plasma, followed by creating of an annular hot-electron plasma confined in a magnetic mirror which consists of two circular electromagnet coils centered on a single axis as is disclosed in U.S. Pat. No. 5,838,760. Continuous emission of bremsstrahlung (X-rays) results from collisions between the highly energetic electrons in the annulus and the background plasma ions and fill gas atoms. It is also contemplated in the present invention that the radiation source may comprise a gamma source. Since gamma sources comprising radionucleotides such as cobalt-60 emit high energy radiation in multiple directions, one or more of the systems described herein may be positioned around the gamma source, permitting the simultaneous radiation processing of a plurality of products. Each system would comprise an adjustable collimator (110), turntable (70), detection system (130), a means for loading and unloading the turntable (e.g. 150), and be individually monitored so that each product receives an optimal radiation dose with a low DUR. In this latter embodiment, one control system (120) may monitor and control the individual components of each system, or the control systems may be used individually. The above description is not intended to limit the claimed invention in any manner, furthermore, the discussed combination of features might not be absolutely necessary for the inventive solution. The present invention will be further illustrated in the following examples. However it is to be understood that these examples are for illustrative purposes only, and should not be used to limit the scope of the present invention in any manner. Radiation Profiles in a Product with Densities of about 0.2 or about 0.8 g/cm3 An accelerator capable of producing an electron beam of 200 kW and 5 MeV is used to, generate X-rays from a tungsten, water cooled converter. The bremsstrahlung energy spectrum of the X-ray beam produced in this manner extends from 0 to about 5 MeV, with a mean energy of about 0.715 MeV. A cylindrical product of 120 cm diameter, comprising a product with an average density of either 0.2 or 0.8 g/cm3 is placed onto a turntable that rotates at least once during the duration of exposure to the radiation beam. The distance from the source plane (converter) to the center of the product is 112 cm. The collimator is set to produce a beam width of 10, 50 or 120 cm. The rectangular cross section of height of the beam is set to the height of the product. Typically to deliver a dose of about 1.5 kGy to a product characterised in having a density of 0.2 g/cm3, the product is exposed to radiation for about 2 to about 2.5 min, while a product having an average density of 0.8 g./cm3 is exposed for about 10 min in order to achieve the desired Dmin. The photon output over the height of the beam was determined for each aperture width, and is constant in both a horizontal and vertical dimension (FIG. 9). Depth dose profiles are determined for three aperture widths, 10, 50 and 120 cm, for a 5 Mev endpoint bremstrahlung x-ray spectrum, with a mean energy of about 0.715 MeV, for each product average density. The results are presented in FIGS. 10(a) and (b)), and Tables 1 and 2. TABLE 1Results for a 0.2 g/cm3 product (see FIG. 10(a))Aperture (cm)DoseMax:DoseMinBeam use efficiency (%)1012.649.5503.148.51201.1441.7 TABLE 2Results for a 0.8 g/cm3 product (see FIG. 10(b))Aperture (cm)DoseMax:DoseMinBeam use efficiency (%)103.188.3501.1687.81203.181.4 Irradiation of Circular and Rectangular Products: 1 mm Convertor Bremsstrahlung X-rays are produced as described above using a 5 MeV electron beam with a circular cross section (10 mm diameter) that scanned vertically across the converter. A 1 mm Ta converter backed with an aluminum (0.5 cm) water (1 cm) aluminum (0.5 cm) cooling channel is used to generate the X-rays. A product of 0.8 g./cm3, with two footprints are tested: one involved a cylindrical product with a 60 cm or 80 cm radius footprint, the other is a rectangular product with a footprint of 100×120 cm, and 180 cm height, both product geometries are rotated at least once during the exposure time. The distance from the converter to the collimator is 32 cm. In order to optimize DUR, several collimator apertures are tested for a cylindrical product (Table 3). Examples of several determinations of the dose along a slice of the product, for a 60 cm radius cylindrical product are presented in FIG. 11. TABLE 3DUR determination for cylindrical products (0.8 g/cm3 density), ofvarying diameter (r), for a range of collimator aperture widths (A)using a 1 cm electron beam producing bremsstrahlung X-rays from a1 mm Ta converter..Dmax:DminAperture, ‘A’ (cm)r = 60r = 70r-8081.631.611.72101.411.381.72111.13nd1.76131.19nd nd151.141.38nd201.381.632.02nd not determined In each tested product diameter, the DUR varied as the collimator aperture changed. Typically, for smaller and larger apertures the DUR is higher when compared with the optimal aperture width. For example, a product of 60 cm diameter exhibites an optimal DUR with a collimator aperture of 11 cm. With this aperture width, the dose is generally uniform throughout the product (see FIG. 11(a)). With an increased width of collimator aperture, of 20 cm, the dose increases towards the periphery of the product, while with a smaller collimator aperture (10 cm), the central portion of the product receives an increase dose (FIG. 11(a)). With a product of increased diameter (80 cm), the DUR increased, and exhibites a greater variation in dose received across the depth of the product (FIG. 11(b)). The general relationship between width of collimator aperture and product diameter, that produces an optimal DUR is shown in FIG. 11(c), where, for a cylindrical product, the lowest DUR is achieved using a narrower aperture with increasing product diameter. For a rectangular product footprint (120 cm×100 cm), the apparent depth of the product, relative to the incident radiation beam, varies as the rectangular product rotates, relative to the beam. In order to optimize the DUR, the collimator aperture width, beam intensity (power), or both, may be dynamically adjusted in order to obtain the most optimal DUR. An example of adjusting aperture width during product rotation is shown in FIG. 12(a). In this example, 8 aperture width adjustments are made over 90° rotation of the product. These same aperture adjustments are mirrored and repeated for the remaining 270° of product rotation so that 32 discrete aperture widths take place during one rotation of a rectangular product. An example of more alterations in aperture width, in this case 26 discrete width in 90° rotation, is shown in FIG. 12(b). However, it is to be understood that the number of discrete aperture widths may vary from the number shown in FIGS. 12(a) and (b), and may include fewer, or more, adjustments as required. For example, for products of lower density, fewer or no adjustments may be required. An optimized DUR may also be obtained through adjustment of the intensity of the radiation beam during rotation of a rectangular product (FIG. 12(c)). In this example, 8 different beam power adjustments are made over 90° rotation of the product. The same beam power adjustments are mirrored and repeated for the remaining 270° rotation of the product. Again, the number of adjustments of beam power, as a function of product rotation, may vary from that shown in order to optimize DUR, depending upon the size and configuration of the product, as well as density of the product itself. In order to further optimize the DUR, both the aperture and beam power may be modulated as the product rotates. When both parameters are modulated, a DUR of from 1.47 to 1.54 was obtained for irradiation of a 0.8 g./cm3, rectangular product (footprint: 120 cm×100 cm), placed at 80 cm from the collimator aperture, using a 1 mm Ta converter (accelerator running at 200 kW, 40 mA electron beam at 5 MeV). Irradiation of Circular and Rectangular Products: 2.35 mm Convertor The Dmax:Dmin ratio may still be further optimized by increasing the overall penetration of the beam within the product. This may be achieved by increasing the thickness of the convertor to produce a X-ray beam with increased average photon energy. In order to balance yield of X-rays and beam energy, a Ta convertor of 2.35 mm (including a cooling channel; 0.5 cm A1, 1 cm H2O, 0.5 cm A1) was selected. This thicker convertor generates fewer photons per beam electron (0.329 phton/beam electron), compared with the 1 mm convertor (0.495 photon/beam electron) due to the increased thickness and attenuation of the X-ray beam. However, even though the number of X-rays produced is lower with a 2.35 mm convertor, the beam that exits the convertor is of a higher average photon energy. As a result of the change in irradiation beam properties, the effect of aperture width and beam power were examined within cylindrical and rectangular products as outlined in Example 2. Results for adjusting the collimator aperture width are presented in Table 4. TABLE 4DUR determination for cylindrical products (0.8 g/cm3 density), ofvarying diameter (r), for a range of collimator aperture widths (A)using a 1 cm electron beam producing bremsstrahlung X-rays from a2.35 mm Ta converter.Dmax:DminAperture, ‘A’ (cm)r = 60r = 70r-808nd1.691.64101.441.431.6 121.281.3 1.64131.32nd141.181.32nd151.14ndnd201.28ndndnd not determined For the irradiation of a rectangular product (120 cm×100 cm; 0.8 g./cm3 density), the collimator aperture may be adjusted to account for changes in the apparent depth of the product relative to the incident radiation beam during product rotation (FIG. 12(b)). As outlined in example 2, the power of the beam may also be adjusted during product rotation (FIG. 12(d)). By adjusting both collimator aperture width and beam power during product rotation, a DUR of from 1.27 to 1.32 is achieved. Irradiation of Circular Product: Effect of Auxiliary Shield The Dmax:Dmin ratio may also be optimized by profiling the beam using an auxiliary shield. Various shapes and types of auxiliary shields were tested (examples of several are shown in FIG. 13(a)). For these analysis, a Ta convertor of 2.35 mm (including a cooling channel; 0.5 cm Al, 1 cm H2O, 0.5 cm Al) is used, with an ebeam energy of 5 Mev (beam current 40 mA; beam power 200 kW max, 78 kW min; 117 kW avg.), an aperture of 9.5 cm., and a distance from the converter to collimator of 32 cm. A circular product (80 cm radius), with a density of 0.8 g/cm3 is tested. Under these conditions, a DUR (Max/Min) value of 1.61 is observed. Results from the insertion of several auxiliary shields (shown in FIG. 13), of varying compositions (Al or Ta) and sizes, within the aperture of the collimator are presented in Table 5. An example of the effect of an auxiliary shield on the dose distribution profiles of a product are shown in FIG. 13(b). The effect of the auxiliary shields on DUR were determined by comparing the Dmin and Dmax values across the entire product diameter (Max/Min 0 to 80 cm), and across the radius (Max/Min 0 to 40). TABLE 5Effect of auxiliary shield on DURAux ShieldMin/MaxMin/MaxtypeMaterialDimension0 to 800 to 40Control——1.611.43A-1Al2.5 cm dia1.631.4 A-2Al 4 cm dia1.631.36B-1Ta2.5 × 0.74 cm21.6 1.37B-2Ta 4 × 1.2 cm21.581.31C-1Ta2.5 cm hr + 1 mm full1.561.36sheetC-2Ta2.5 cm hr* + 2 mm full1.521.35sheetC-3Ta2.5 cm hr* + 3 mm full1.511.36sheetDTa 3 mm full sheet1.531.51*hr - half-rod As can be seen from Table 5, the use of Ta as an auxiliary shield reduced the DUR (both Max/Min 0 to 80, and 0 to 40). Furthermore, the shape and size of the shield may be varied to further optimize the DUR within a product. In the absence of an auxiliary shield, the overall dose received by the product was higher than that observed in the presence of a shield (FIG. 13(b)), and characterized as having a higher dose received in the outer regions of the product, and reduce dose in the central region. In the presence of the auxiliary shield, even though the central region received a lower dose, thereby reducing the difference between Dmax and Dmin (lower DUR), the outer regions of the product also received a lower dose. The dose distribution profile obtained in the presence of an auxiliary shield was in general characterized as having reduced the overall radiation dose received, and by producing a flatter dose distribution profile throughout the product. The improved results are obtained using an auxiliary shield that spanned the entire collimator aperture, thereby only permitting X-rays of higher energy to enter the product (i.e. hardened the X-ray spectrum). Irradiation of Circular Product: Effect of Beam Offset The Dmax:Dmin ratio may also be optimized by offsetting the beam from the axis of product rotation so that the relative fractional exposure time within the different lateral parts of the product are altered. For these analyses, a Ta convertor of 2.35 mm (including a cooling channel; 0.5 cm Al, 1 cm H2O, 0.5 cm Al) is used, with an ebeam energy of 5 Mev (beam current 40 mA; beam power 200 kW max, 78 kW min; 117 kW avg.), an aperture of 9.5 cm., and a distance from the converter to collimator of 32 cm. A rectangular product (100×120 cm), with a density of 0.8 g/cm3 is tested. During radiation, the collimator aperture is modified (as described in Example 2) during rotation of the rectangular product from a min value of 11.5 cm to a max value of 17.5 cm (FIG. 14(a)). Also, the beam power is modified as shown in FIGS. 14(b) respectively (also see Example 3). In the present example, beam offset of 7 cm, with respect to the product center, is tested. A beam offset of 7 cm is obtained by angling the beam (aperture inclination angle, ΘA), by 5° from the center line of the beam. Under these conditions, a DUR (Max/Min) value of 1.4 is observed (FIG. 14(c)). However, the use of a narrower collimator aperture (less than 11.5 cm) further reduces the higher doses received at the periphery of the product, and produces a DUR of 1.2. The dose distribution profile produced as a result of the beam offset is characterized as having smaller regions of low dose, with a higher uniformity across the product. All publications are herein incorporated by reference. The present invention has been described with regard to preferred embodiments. However, it will be obvious to persons skilled in the art that a number of variations and modifications can be made without departing from the scope of the invention as described herein.
048333298	description	DETAILED DESCRIPTION Referring to the drawings, first more particularly to FIG. 4, a system of this invention for generating and containerizing radioisotopes is shown to comprise a generator 1 containing a sterile pyrogen-free supply indicated at 3 of a parent radioisotope. As disclosed in the aforesaid U.S. Pat. No. 4,296,785, generally, this generator comprises an elongate cylindric glass tube having piercable closures indicated at 5 and 7 at its upper and lower ends (upper and lower as shown in FIG. 4) each constituted by a rubber stopper plugged in the respective end of the container. An aluminum crimp cap 9 is shown for each stopper with a central section of the cover removed. The parent radioisotope may be molybdenum-99, for example, adsorbed on an anion exchange medium, alumina or other medium (as in U.S. Pat. No. 3,655,981) for generating technetium-99M. The generator could be a tin/indium or germanium/gallium generator. Pierced through the rubber stopper at the upper end of the generator is the downturned end 11 of a relatively thin metal tube 13 constituting an eluant inlet for the generator. Pierced through the rubber stopper at the lower end of the generator is the upturned end 15 of a relatively thin metal tube 17 constituting an eluant outlet for the generator. At 19 is shown a reservoir for holding a supply of eluant indicated at 21 (e.g. saline solution). Preferably, this is a glass bottle having a rubber stopper 23 in its mouth with an aluminum foil cover 25 for the stopper, shown in FIG. 4 with a central circular section of the cover removed. Pierced through the stopper 23 is the downturned end 27 of a relatively thin metal tube 29 constituting an air inlet for the reservoir or bottle, for admission of air to the bottle to apply atmospheric air pressure on the eluant 21 in the bottle. Also pierced through the stopper 23 is the downturned leg 31 of a thin metal tube 33 constituting an eluant outlet for the bottle. The downturned leg 31 of the tube 33 extends down in the bottle nearly to the bottle of the bottle for the delivery of eluant upwardly through the leg 31. At 35 is indicated a tubular needle for piercing the rubber stopper 37 of a sealed sterile evacuated container or vial 39 (which may be placed in a lead shield as indicated at 40 in FIGS. 1 and 2). The tubular needle 35 extends down from the lower end of the valve body 41 of valve means of this invention which is designated in its entirety by the reference numeral 43. The valve body 41 is carried by a case indicated generally at 45 in FIGS. 1 and 2 corresponding generally to the case shown in U.S. Pat. No. 3,655,981, for movement downwardly from its raised retracted position of FIGS. 1 and 2 against the upward bias of a return spring 47 for causing the needle to pierce the rubber stopper or closure 37 of the vial 39, and for movement back upwardly to its raised retracted position by the spring. The case is shown as including an overhanging portion having top and bottom walls 49 and 51 and an outer wall 53 for mounting the body 41. A first flexible tube 55, constituted of a length of plastic tubing, which is resiliently compressible and thereby adapted to be pinched for closing it, is suitably connected at one end to the air inlet tube 29 for the eluant bottle or reservoir 19 and is in communication at its other end with the atmosphere upstream from the reservoir via a bacteriological filter 56 for precluding entry of bacteria from the atmosphere to the system. This tube is adapted to be pinched to close it, as will appear, to block communication between the head space 57 in the eluant reservoir above the eluant therein and the atmosphere. When the tube is open, the head space 57 is in communication with the atmosphere for subjecting the eluant in the eluant reservoir to atmospheric pressure for flow of eluant from the reservoir via the eluant outlet tube 33. A second flexible tube designated 59, also constituted by a length of resiliently compressible pinchable plastic tubing, is suitably interconnected between the eluant outlet tube 33 and the generator inlet tube 13. This tube 59 is adapted to be pinched closed, as will appear, to block communication between the eluant reservoir and the generator. When tube 59 is open, eluant may flow from the reservoir to the generator. A third flexible tube designated 61, also constituted by a length of resiliently compressible pinchable plastic tubing, is interconnected between the generator outlet tube 17 and the tubular needle 35. This tube is adapted to be pinched closed, as will appear, to block communication between the generator and the tubular needle. When the tube 61 is open, eluate may flow from the generator to the needle. Each of the tubes 55, 59, and 61 may be tubing made of plastic such as that sold under the trade name Silastic by Dow Corning Corp. of Midland, Michigan, of 0.156" outside diameter and 0.036" inside diameter. The tubes 13, 17, 29, 33 may be 19 gauge stainless steel tubes with beveled ends for piercing the respective stoppers and the connections to the tubes 13 and 17 may be by female luer fittings such as indicated at 63. The valve means 43 is provided for pinching the three tubes 55, 59 and 61 to close them and is operable on entry of the tubular needle 35 through the closure 37 of an evacuated container of vial 39 to open the tubes for venting of the eluant reservoir 19 to atmosphere via tube 55, for delivery of eluant from the reservoir to the generator via the tube 59, and for delivery of eluate from the generator via tube 61 to the needle 35 and thence to the evacuated container or vial 39. The body 41 of the valve means may be molded in one piece of a suitable plastic, such as polypropylene, with a generally elongate stem 65 which extends vertically as used in the system and which thereby has an upper end at 67 and a lower end at 69, and a head 71 at the upper end of the stem. The stem 65 is generally tubular so as to have an axial passage 73 for the tube 61. At its lower end 69 the stem is formed with an enlarged socket having an internal diameter somewhat larger than that of the passage 73, this socket having an internal annular groove 75. The passage 73 extends all the way down in the stem from its upper end to the socket, opening at its lower end into the socket. The stem is formed with an axial slot 77, flanges 79 being provided at opposite sides of the axial slot for stiffening the stem. The head 71 has a bottom 81 and spaced side walls each generally designated 83 extending up from the bottom, the side walls 83 defining a recess 85 in the head which is open at the top of the head. The side walls have portions 87 extending parallel to one another from a transverse wall 89 of the head, which may be referred to as the back wall of the head, and forward portions which converge to a relatively narrow wall 93 which may be referred to as the front wall of the head. Each of the forward portions 91 of the side walls 83 of the head has a slot 95 therein extending down from the top edge of the respective wall adjacent the rearward ends of said forward portions generally in a vertical plane parallel to and spaced from the narrow front wall 93 of the head. The slots have a width slightly greater than the diameter of the first, second and third tubes 55, 59 and 61, which are received in these slots one on top of another, tube 61 being the first to be placed in the slots and hence the lowest of the three tubes, tube 59 being next and hence being the intermediate tube in the slots, and tube 55 being the last to be placed in the slots and hence being the uppermost of the three tubes. Portions of the tubes, so lodged in the slots, extend across the head across the rearward edge of a back-up member 97 extending rearward from the front wall 93 of the head, being bent to some extent around the rearward edge of this back-up member. The latter is constituted by a vertical rib formed integrally with the head extending rearward from the narrow front wall 93 of the head slightly beyond the plane of the forward edges of the slots 95. The stated portions of the tubes also extend over vertical ribs 99 extending between wall 89 of the head and forward portions 91 of the head side walls 83 and a central vertical rib 101. The portions of the tubes 55, 59 and 61 which extend across the head 71 are adapted to be pinched or clamped closed against the rearward edge of the back-up member 97 by means comprising a pinch member 103 mounted for swinging movement on one of the side walls 83 of the head extending toward the other side wall above the ribs 99 and 101 and cam means 107 rotary in the head for swinging the pinch member into engagement with the portions of the tubes extending across the head to pinch or clamp them against the back-up member 97, and for releasing the pinch member to allow it to swing away from said portions of the tubes to allow them to open up. The pinch member is constituted by a flexible molded plastic plate member of generally rectagular shape having a flange 109 at one edge with a thickened bead 111 at the edge of the flange. The pinch plate 101 is mounted at said one side wall of the head by forming of that side wall with a vertical groove 113 on the inside and a flange 115 extending inwardly from that side wall at the rear of the groove with a lip 117 at the inner edge of the flange extending forward over the groove. The pinch plate is assembled with the head by sliding the bead 111 down into the recess defined by the groove 113, the flange 115 and the lip 117. The plate is adapted to flex adjacent the flange for the swinging of the plate generally on what is in effect an integral hinge adjacent its end with the bead 111. The rotary cam means 107 comprises a molded plastic (e.g. polypropylene) member formed to have a circular disk constituting a cap 119 for engagement with the top of the head 71, a cam shaft 121 extending down from the disk with a lug 123 extending radially outwardly from this shaft constituting the cam proper, the outer edge of the lug being rounded. The lug 123 extends down from the disk or cap 119 toward but terminates short of the lower end of the shaft 121. The latter is made tubular and is formed on its exterior with an annular snap ring formation or rib 125 located below the lower end of the lug 115. The lower end portion of the shaft, below the lower end of the lug, is rotatably fitted in a generally cylindric bearing 127 formed in the head, this bearing having an annular groove 129 in which the annular snap ring formation or rib 125 on the shaft is snap fitted. A hub 131 extends up from the disk 119 coaxial with the shaft 121, this hub being tubular as appears in FIG. 4 and having an axial external key 133 extending throughout its length. A knob or handle 135 has a stem 137 fitted in the hub 131, the stem 137 having a keyway 139 which interfits with a key 141 in the hub for keying the knob to the hub. Stem 137 also has a stop 143 engageable with the upper end of the hub. With the parts 41 and 107 disassembled, the upper end of the head 43 is open for insertion of the plate 103 in the head, also for insertion of the tubes 61, 59 and 55 (in that order) in the slots 95 in the side walls 83 of the head. The rotary member 107 may then be snap-fitted into assembly with the head. This makes for easy assebly. With member 107 so assembled with the head, the pinch plate 103 and the three tubes are held in place by the disk or cap 119 of member 107. The eluate tube 61 is connected at one end to the eluate outlet tube 17 for the generator 1. It extends in a loop from the head 71 of the valve body 41 and through the slot 77 in the stem into the passage 73 in the stem 65 and down in the passage 73 to a fitting 145, e.g. a luer fitting, accommodated in the socket 69 at the lower end of the stem to which the needle 35 is removably attached. The fitting 145 has an annular rib 147 snapped into the groove 75. The cam member 107 normally occupies the tubepinching position in which it is illustrated in FIG. 5 herein the cam lug 123 is in line with the back-up rib 97 and holds the pinch plate 103 against the tubes 55, 59 and 61 to pinch the tubes closed (see also FIG. 3). This tubepinching position of the cam member is generally determined by engagement of the lug 123 with the edge 149 of an upwardly extending portion of the bearing 127 in the head 71. The valve body 41 with the needle 35 extending down from its lower end, with the tubes 55, 59 and 61 in the slots 95, with the cam member 107 in place in the head 71 and in its tube-pinching position, and with knob 135 on the cam member, is movable downwardly against the bias of the spring 47 to drive the needle through the rubber stopper 37 of an evauated vial 39. The knob is then turned to turn the cam member to its tube-release position of FIG. 6 for opening the tubes for the flow of eluant from the bottle 19 via tube 59 to the generator 1 and for flow of eluate from the generator 1 via tube 61 to the needle 35 and thence into the vial. On delivery of the requisite amount of eluate into the vial, the knob is turned back to return the cam member 107 to its FIG. 5 tube-pinching position to cut off flow, and the valve body (with the needle and tubes) is released for return upward to its raised retracted position of FIGS. 1 and 2 awaiting filling of the next vial. The pinching of the tubes by the hinged pinch plate 103 is such as to avoid displacement ("walking") and stretching of the tubes, and also to avoid abrasion of the tubes. In view of the above, it will be seen that the several objects of the invention are achieved and other advantageous results attained. As various changes could be made in the above constructions without departing from the scope of the invention, it is intended that all matter contained in the above description or shown in the accompanying drawings shall be interpreted as illustrative and not in a limiting sense.
	claims	1. A stage mechanism, comprising: a first stage movable along a reference plane, containing a vertical direction, wherein said first stage is movable in a first direction corresponding to one of the vertical direction and a direction approximating the vertical direction; a second stage movable in a second direction intersecting with the first direction and relative to said first stage; a first driving mechanism for moving said first stage in the first direction; a second driving mechanism for moving said second stage in the second direction; a countermass movable in the first direction, said countermass having a mass balancing with a mass of one of said first and second stages, which moves in the first direction; and a third driving mechanism for generating a force for moving said countermass in a direction opposite to the first direction. 2. A stage mechanism according to claim 1 , further comprising a secondary countermass and a fourth driving mechanism for moving the secondary countermass in a direction opposite to the second direction. claim 1 3. A stage mechanism according to claim 1 , wherein each of said first and second driving mechanisms comprises a linear motor. claim 1 4. A stage mechanism according to claim 2 , wherein each of said third and fourth driving mechanisms comprises a linear motor. claim 2 5. A stage mechanism according to claim 1 , wherein said stage mechanism includes plural countermasses each being said countermass. claim 1 6. A stage mechanism according to claim 1 , wherein said third driving mechanism is controlled so that a reaction force or a change in gravity center produced due to the motion of the first stage is reduced by movement of the countermass. claim 1 7. A stage mechanism according to claim 2 , wherein said fourth driving mechanism is controlled so that a reaction force or a change in gravity center produced due to the motion of the second stage is reduced by movement of the countermass. claim 2 8. A stage mechanism according to claim 1 , further comprising compensating means for compensating for a weight of the first stage and the countermass. claim 1 9. A stage mechanism according to claim 8 , wherein said compensating means has a connecting mechanism for connecting the stage and the countermass. claim 8 10. A stage mechanism according to claim 9 , wherein said connecting mechanism includes a pulley and a belt. claim 9 11. A stage mechanism according to claim 10 , wherein said pulley has a pulley diameter ratio substantially corresponding to an inverse of a mass ratio between the first stage and the countermass. claim 10 12. A stage mechanism according to claim 9 , wherein said connecting mechanism includes a cylinder mechanism. claim 9 13. A stage mechanism according to claim 12 , wherein said cylinder mechanism is arranged so that the weight of the first stage and the weight of the countermass are balanced with each other. claim 12 14. A stage mechanism according to claim 13 , wherein a sectional area of a piston for supporting the first stage and a sectional area of a piston for supporting the weight of the countermass substantially correspond to a mass ratio between the first stage and the countermass. claim 13 15. A stage mechanism according to claim 1 , further comprising a damper for supporting said stage system. claim 1 16. A stage mechanism, comprising: a first stage movable along a reference plane, containing a vertical direction, and in the vertical direction or in a first direction close to the vertical direction; a second stage movable in a second direction intersecting with the first direction and relative to the first stage; a countermass movable in the first direction; a first driving mechanism for moving the first stage in the first direction; and a second driving mechanism for moving the second stage in the second direction, wherein said first driving mechanism includes a magnet and a coil, one of which is mounted on one of the first stage and the countermass and the other of which is mounted on the other of the first stage and the countermass, such that the countermass moves in a direction opposite to the first stage. 17. A stage mechanism according to claim 16 , further comprising a secondary countermass movable in a direction opposite to the second direction. claim 16 18. A stage mechanism according to claim 17 , wherein said second driving mechanism includes a magnet and a coil, one of which is mounted on one of the second stage and the secondary countermass and the other of which is mounted on the other of the second stage and the secondary countermass, such that the secondary countermass moves in a direction opposite to the second stage. claim 17 19. A stage mechanism according to claim 16 , further comprising compensating means for compensating for a weight of the first stage and the countermass. claim 16 20. A stage mechanism according to claim 19 , wherein said compensating means has a connecting mechanism for connecting the stage and the countermass. claim 19 21. A stage mechanism according to claim 20 , wherein said connecting mechanism includes a pulley and a belt. claim 20 22. A stage mechanism according to claim 21 , wherein said pulley has a pulley diameter ratio substantially corresponding to an inverse of a mass ratio between the first stage and the countermass. claim 21 23. A stage mechanism according to claim 20 , wherein said connecting mechanism includes a cylinder mechanism. claim 20 24. A stage mechanism according to claim 23 , wherein said cylinder mechanism is arranged so that the weight of the first stage and the weight of the countermass are balanced with each other. claim 23 25. A stage mechanism according to claim 24 , wherein a sectional area of a piston for supporting the first stage and a sectional area of a piston for supporting the weight of the countermass substantially correspond to a mass ratio between the first stage and the countermass. claim 24 26. A stage system, comprising: a stage movable along a reference plane, containing a vertical direction, and in the vertical direction or a direction close to the vertical direction; a countermass, balancing with a weight of the stage; connecting members for connecting the countermass to the stage; a pulley for supporting said connecting members; and an anti-vibration mechanism for reducing vibration to be propagated from said connecting members to the stage. 27. A stage mechanism according to claim 26 , wherein said anti-vibration mechanism is provided at one of a connection between the connecting members and the countermass, and a connection between the stage and the connecting members. claim 26 28. A stage system according to claim 27 , wherein said anti-vibration mechanism includes an elastic member. claim 27 29. A stage system according to claim 28 , wherein said elastic member comprises laminated rubber. claim 28 30. A stage system according to claim 26 , wherein said anti-vibration mechanism includes an elastic member provided at one of a supporting portion for supporting a bearing of the pulley and the surface of the pulley. claim 26 31. A stage system according to claim 26 , wherein said anti-vibration mechanism comprises a rotary type static bearing for rotatably supporting a bearing of the pulley. claim 26 32. A stage system according to claim 26 , wherein said anti-vibration mechanism includes an actuator for producing a force. claim 26 33. A stage system according to claim 26 , further comprising a static bearing device for maintaining the stage out of contact with the reference plane. claim 26 34. A stage system according to claim 26 , wherein the stage comprises a two-dimensional stage movable two-dimensionally along the reference plane. claim 26 35. A stage system according to claim 34 , further comprising a yaw guide for guiding the stage in the vertical direction, and a yaw guide static bearing for maintaining the stage out of contact with the yaw stage. claim 34 36. A stage system according to claim 26 , wherein each of the connecting members comprises one of a steel belt and a steel wire. claim 26 37. A stage system, comprising: a stage movable along a reference plane, containing a vertical direction, and in the vertical direction or a first direction close to the vertical direction; connecting members connected to the stage; a pulley for supporting said connecting members; a motor for rotationally driving the pulley and for compensating for a weight of the stage; and an anti-vibration mechanism for reducing vibration to be propagated from said connecting members to the stage. 38. A stage system according to claim 37 , wherein said anti-vibration mechanism is provided at a connection between the stage and the connecting members. claim 37 39. A stage system according to claim 38 , wherein said anti-vibration mechanism includes an elastic member. claim 38 40. A stage system according to claim 39 , wherein said elastic member comprises laminated rubber. claim 39 41. A stage system according to claim 37 , wherein said anti-vibration mechanism includes an elastic member provided at one of a supporting portion for supporting a bearing of the pulley and the surface of the pulley. claim 37 42. A stage system according to claim 37 , wherein said anti-vibration mechanism comprises a rotary type static bearing for rotatably supporting a bearing of the pulley. claim 37 43. A stage system according to claim 37 , wherein said anti-vibration mechanism includes an actuator for producing a force. claim 37 44. A stage system according to claim 37 , further comprising a static bearing device for maintaining the stage out of contact with the reference plane. claim 37 45. A stage system according to claim 37 , wherein the stage comprises a two-dimensional stage movable two-dimensionally along the reference plane. claim 37 46. A stage system according to claim 45 , further comprising a yaw guide for guiding the stage in the vertical direction, and a yaw guide static bearing for maintaining the stage out of contact with the yaw guide. claim 45 47. A stage system according to claim 37 , wherein each of the connecting members comprises one of a steel belt and a steel wire. claim 37 48. A stage system, comprising: a stage movable along a reference plane, containing a vertical direction, and in the vertical direction or a direction close to the vertical direction; a countermass, balancing with a weight of the stage; connecting members for connecting the countermass to the stage; a pulley for supporting said connecting members; and an actuator for adjusting one of a tension force of and an effective length of the connecting members. 49. A stage system according to claim 48 , further comprising control means for controlling said actuator on the basis of positional information about the stage. claim 48 50. A stage system according to claim 48 , wherein said actuator is provided at one of a connection between the connecting members and the countermass, and a connection between the stage and the connecting members. claim 48 51. A stage system according to claim 48 , wherein said actuator is provided at a supporting portion for supporting the pulley. claim 48 52. A stage system according to claim 48 , wherein said actuator includes at least one of an air spring, an air cylinder, a linear motor and a piezoelectric device. claim 48 53. A stage system according to claim 48 , further comprising a static bearing device for maintaining the stage out of contact with the reference plane. claim 48 54. A stage system according to claim 48 , wherein the stage comprises a two-dimensional stage movable two-dimensionally along the reference plane. claim 48 55. A stage system according to claim 54 , further comprising a yaw guide for guiding the stage in the vertical direction, and a yaw guide static bearing for maintaining the stage out of contact with the yaw guide. claim 54 56. A stage system according to claim 48 , wherein each of the connecting members comprises one of a steel belt and a steel wire. claim 48 57. An exposure apparatus, comprising: a first stage movable along a reference plane, containing a vertical direction, wherein the first stage is movable in a first direction corresponding to one of the vertical direction and a direction approximating the vertical direction; a second stage movable in a second direction intersecting with the first direction and relative to the first stage; a first driving mechanism for moving the first stage in the first direction; a second driving mechanism for moving the second stage in the second direction; a countermass movable in the first direction, said countermass having a mass balancing with a mass of one of said first and second stages, which moves in the first direction; and a third driving mechanism for generating a force for moving the countermass in a direction opposite to the first direction. 58. An apparatus according to claim 57 , wherein said exposure apparatus comprises an X-ray exposure apparatus. claim 57 59. An exposure apparatus, comprising: a first stage movable along a reference plane, containing a vertical direction, and in the vertical direction or in a first direction close to the vertical direction; a second stage movable in a second direction intersecting with the first direction and relative to the first stage; a countermass movable in the first direction; a first driving mechanism for moving the first stage in the first direction; and a second driving mechanism for moving the second stage in the second direction, wherein said first driving mechanism includes a magnet and a coil, one of which is mounted on one of the first stage and the countermass and the other of which is mounted on the other of the first stage and the countermass, such that the countermass moves in a direction opposite to the first stage. 60. An apparatus according to claim 59 , wherein said exposure apparatus comprises an X-ray exposure apparatus. claim 59 61. An exposure apparatus, comprising: a stage movable along a reference plane, containing a vertical direction, and in the vertical direction or a direction close to the vertical direction; a countermass, balancing with a weight of the stage; connecting members for connecting the countermass to the stage; a pulley for supporting said connecting members; and an anti-vibration mechanism for reducing vibration to be propagated from said connecting members to the stage. 62. An apparatus according to claim 61 , wherein said exposure apparatus comprises an X-ray exposure apparatus. claim 61 63. An exposure apparatus, comprising: a stage movable along a reference plane, containing a vertical direction, and in the vertical direction or a first direction close to the vertical direction; connecting members connected to the stage; a pulley for supporting said connecting members; a motor for rotationally driving the pulley and for compensating for a weight of the stage; and an anti-vibration mechanism for reducing vibration to be propagated from said connecting members to the stage. 64. An apparatus according to claim 63 , wherein said exposure apparatus comprises an X-ray exposure apparatus. claim 63 65. An exposure apparatus, comprising: a stage movable along a reference plane, containing a vertical direction, and in the vertical direction or a direction close to the vertical direction; a countermass, balancing with a weight of the stage; connecting members for connecting the countermass to the stage; a pulley for supporting said connecting members; and an actuator for adjusting one of a tension force of and an effective length of the connecting members. 66. An apparatus according to claim 65 , wherein said exposure apparatus comprises an X-ray exposure apparatus. claim 65 67. A device manufacturing method, comprising the steps of: applying a photosensitive material to a wafer; exposing the wafer by use of an exposure apparatus; and developing the exposed wafer, wherein the exposure apparatus comprises (i) a first stage movable along a reference plane, containing a vertical direction, wherein the first stage is movable in a first direction corresponding to one of the vertical direction and a direction approximating the vertical direction, (ii) a second stage movable in a second direction intersecting with the first direction and relative to the first stage, (iii) a first driving mechanism for moving the first stage in the first direction, (iv) a second driving mechanism for moving the second stage in the second direction, (v) a countermass movable in the first direction, the countermass having a mass balancing with a mass of one of the first and second stages, which moves in the first direction, and (vi) a third driving mechanism for generating a force for moving the countermass in a direction opposite to the first direction. 68. A device manufacturing method, comprising the steps of: applying a photosensitive material to a wafer; exposing the wafer by use of an exposure apparatus; and developing the exposed wafer, wherein the exposure apparatus comprises (i) a stage movable along a reference plane, containing a vertical direction, and in the vertical direction or a direction close to the vertical direction, (ii) a countermass, balancing with a weight of the stage, (iii) connecting members for connecting the countermass to the stage, (iv) a pulley for supporting the connecting members, and (v) an anti-vibration mechanism for reducing vibration to be propagated from the connecting members to the stage. 69. A device manufacturing method, comprising the steps of: applying a photosensitive material to a wafer; exposing the wafer by use of an exposure apparatus; and developing the exposed wafer, wherein the exposure apparatus comprises (i) a stage movable along a reference plane, containing a vertical direction, and in the vertical direction or a first direction close to the vertical direction, (ii) connecting members connected to the stage, (iii) a pulley for supporting the connecting members, (iv) a motor for rotationally driving the pulley and for compensating for a weight of the stage, and (v) an anti-vibration mechanism for reducing vibration to be propagated from the connecting members to the stage. 70. A device manufacturing method, comprising the steps of: applying a photosensitive material to a wafer; exposing the wafer by use of an exposure apparatus; and developing the exposed wafer, wherein the exposure apparatus comprises (i) a stage movable along a reference plane, containing a vertical direction, and in the vertical direction or a direction close to the vertical direction, (ii) a countermass, balancing with a weight of the stage, (iii) connecting members for connecting the countermass to the stage, (iv) a pulley for supporting the connecting members, and (v) an actuator for adjusting one of a tension force of and an effective length of the connection members. 71. A stage mechanism, comprising: a first stage movable along a reference plane, containing a vertical direction, and in the vertical direction or in a first direction close to the vertical direction; a second stage movable in a second direction intersecting with the first direction and relative to said first stage; a first driving mechanism for moving said first stage in the first direction; a second driving mechanism for moving said second stage in the second direction; a countermass movable in the first direction; a third driving mechanism for moving said countermass in a direction opposite to the first direction; and compensating means for compensating for a weight of said first stage and said countermass, wherein said compensating means comprises a connecting mechanism for connecting said stage and said countermass, said connecting mechanism includes a pulley and a belt, and said pulley has a pulley diameter ratio substantially corresponding to an inverse of a mass ratio between said first stage and said countermass.
049884751	description	DESCRIPTION OF PREFERRED EMBODIMENT FIG. 1 shows a fuel assembly which comprises a framework shown in FIG. 2 and consisting of spacer grids 2 spaced in the longitudinal direction of the assembly, guide tubes 3 to which the grids 2 are rigidly fixed, an upper joining piece 4 and a lower joining piece 5 which are fixed to the fuel rods 6 of the assembly, which are shorter. The fuel pencils or fuel rods of the assembly which have a than the guide tubes 3, are disposed in the framework so as to form a bundle in which the rods are disposed parallel to one another. The rods are held laterally by the spacer grids 2 so as to form a uniform network with squared mesh, in the transverse sections of the assembly. The spacer grids 2 also ensure the retention of the fuel rods 6 in their longitudinal direction by virtue of gripping means which will be described with reference to FIGS. 3 and 4. FIG. 3 shows a spacer grid 2 consisting of an assembly of small metal plates 8 disposed and assembled at a right angle so as to form a network with squared mesh, the cells 9 of each of which can receive a fuel rod 6 in order to ensure its positioning in the network, its lateral retention and its longitudinal retention. Certain cells 7 are intended to receive a guide tube 3 which is fixed rigidly on the walls of the cell 7. The spacer grids 2 and the guide tubes 3 thus form a rigid framework which is capable of receiving the fuel rods 6 which can be slid, via one end of the assembly, into an assembly of cells 9 in alignment in the longitudinal direction of the framework, as shown in FIG. 2. Each of the cells 9 intended to receive a fuel rod comprises bosses 15 projecting inwards in the cell 9 and springs 16 generally comprising two active parts in two adjacent cells 9 of the spacer grid. In each of the cells 9, two walls at 90.degree. each comprise two bosses 15 projecting inwards in the cell 9 and the two other walls at 90.degree. each comprise a convex part projecting inwards in the cell 9 and in an intermediate position between the bosses 15 in the height of the spacer grid 2. Each of the rods 6 of the assembly is thus in contact with four bosses and two springs inside each of the cells 9. The fuel rod 6 is thus retained by contact of the sheath of this rod at six points with the gripping means of the cell. In the embodiment of a spacer grid shown in FIGS. 3 and 4, the springs 16 are made of a nickel alloy and are attached to the small plates 8 made from zirconium alloy. The bosses 15 are produced by deforming the metal of the small plates 8. The spacer grid 2 is delimited, at its outer contour whose square form corresponds to the section of the fuel assembly, by a frame consisting of small plates 10 assembled at a right angle at the corners of the spacer grid 2. As may be seen in FIGS. 3 and 4, the upper edge of each of the small plates 10 is cut in order to form successive fins 12 which are folded towards the inside of the spacer grid in order to form, with the plane of the small plate 10, an angle .alpha. of well-defined value. The lower edge of each of the small plates 10 of the frame of a spacer 2 of the assembly also comprises guide fins 13 folded towards the inside of the spacer. The spacer grids also comprise, in the extension of the upper edge of each of the cells 9 receiving a fuel rod, fins 14, called mixing fins, used for mixing the cooling water circulating in contact with the rod where it emerges from the spacer grid. A homogenization of the temperature of the water removing the heat supplied by the fuel rods of the assemblies is thus obtained. The small plates 10 of the frame of the spacer grid 2 also comprise two bosses 17 at the level of each of the cells 9, which project inwards in the corresponding cell. The fuel rods disposed in rows adjacent to the small plates 10 of the frame of the spacer grid, such as the rod 6a, and referred to as peripheral rods, are held by two springs 16, two bosses 15 and two bosses 17, these rods being held in each of the cells 9 by contact at six points, like the other rods of the fuel assembly. A sleeve 18, by means of which fastening of the guide tube 3 is ensured, is fixed in each of the cells 7 intended to receive a guide tube 3. The metal forming the support bosses 15 and 17 and the retention springs 16 undergoes, in the environment of the reactor, under irradiation, transformations which can result in a loss of hardness or of resilience. In certain cases, the axial retention force on the pencil is no longer sufficient to ensure its support and the pencil is displaced downwards inside the framework. The peripheral rods 6a disposed in the cells 9 adjacent to the plates 10 of the frame of the spacer grid are particularly stressed and may be displaced longitudinally or subjected to vibrations which cause their breakage. It is thus desirable to be able to check that the retention forces on the peripheral rods 6a of the assembly have retained sufficient values before refueling, in the core of the reactor, of a spent assembly. The device according to the invention, shown in FIG. 5 and denoted generally by the reference 20, makes it possible to check the axial retention forces on the peripheral rods of an assembly 1 disposed inside a fuel assembly storage pool. The fuel assembly 1 rests, by means of its lower joining piece 5, on a support 21, the positioning of the assembly in a vertical checking position being facilitated by an insertion device which is flared towards the top 22, fixed on the base 21 and which receives the lower part of the assembly. The upper part of the assembly is located at a depth, below the upper level 27 of the pool, substantially equal to three meters, this depth of water ensuring effective biological protection for an operator 23 conducting the checking operations from the platform 26 of a control station 25 fixed on the top part 24 of a lateral wall of the pool. The device according to the invention comprises a rod 28 of great length disposed vertically in the vicinity of the wall 24 of the pool and connected at its upper part 29 to a vertical support which is integrally attached to the platform 26. A carriage 30 is mounted so as to be movable in the longitudinal direction of the rod 28 and may be displaced in this longitudinal direction by means of a winch 31 which may be activated by a handle used by the operator 23. The carriage 30 carries, by means of a displacement device 32, a support 33 on which is fixed the pushing device of the checking assembly 20 and at least one video camera 34. The insertion device 22 is unable to interfere with the positioning and implementation of the device for pushing on the rod 6a. The pushing device is remotely controlled by remote control means 35 of known type, such as ball remote controls, comprising a sheath inside which a flexible element is mounted so as to be movable and which makes it possible to provide a pulling or pushing action which is remotely controlled, for example manually. The ends of the ball remote controls 35 opposite to the support 33 of the pushing device are connected, in the manner which will be described hereinbelow, to control and measurement devices fixed on the railing 37 of the platform 26 of the control station 25. Ball remote controls 36 also make it possible to orient the video cameras, such as 34, so as to provide the operator 23 in the control station 25 with a video image of the zone in which checking is being performed on the retention force on a peripheral rod of the assembly 1. In FIG. 5, the carriage 30 and the displacement device 32 are in a position which makes it possible to push on the upper end of a peripheral rod 6a of the assembly in the vicinity of its upper joining piece 4. The device for the support and positioning of the pushing device is the subject of a patent application filed jointly by the company FRAMATOME and the company COGEMA on the same day as the present patent application. FIGS. 6 to 8 show the displacement device 32 of the support 33 which consists of a crossed carriage displacement assembly. This displacement assembly is fixed on the carriage 30 so as to be movable in the vertical direction and comprises a first carriage, or lower carriage, which is movable in a first horizontal direction substantially perpendicular to the wall 24 of the pool, on which a second carriage, or upper carriage, carrying the support device 33 is mounted so as to be movable in a second horizontal direction substantially parallel to the wall of the pool 24. The vertical displacement carriage 30 and the crossed carriage assembly 32 makes it possible to place the support 33 and the pushing device carried by this support above the upper end of any peripheral rod of the assembly or below the lower end of each of the peripheral rods. The displacements of the crossed carriage assembly 32 can be ensured by ball remote controls, such as the remote controls 35 and 36, whose upper part is accessible from the control station 25. A video image of the work zone provided by the camera 34 enables the operator to place the support 33 and the pushing device in the desired position. A description will now be given, with reference to FIGS. 6, 7 and 8, of the pushing device carried by the support 33. The support 33 consists of two support bases 38 on which are fixed, by means of accurate positioning adjusters 39, brackets for fixing the various elements of the pushing device and the video cameras 34 and 37. Two brackets 40 and 41 connected together by means of fixing screws and nuts support the end of the sheath 42 of a ball remote control comprising a threaded part 44 engaged in an opening provided in the end of the bracket 41. The end of the sheath is fixed by nuts 43 engaged on the threaded part 44 on either side of the bracket 41. Inside the sheath 42, the ball remote control comprises a flexible element 45 mounted so as to move in the axial direction by virtue of balls ensuring low-friction guiding. The flexible element 45 is connected, by means of a joining piece 46, to the end of a rack 47 controlling the pushing device of the fuel rods which will be described hereinbelow. The end of the flexible cable 45 is rigidly fixed on the joining piece 46 by means of a nut 48. The joining piece 46 comprises a groove 49 in which a profiled a part of corresponding form machined at the end of the rack 47 may be engaged. A link is thus provided between the flexible element 45 and the rack 47 which makes it possible to transmit the pushing or pulling action of the flexible element 45 to the rack 47. As may be seen in FIG. 7A, screws 50 make it possible to fix the rack 47 on the fixing piece 46. This figure also shows that the bracket 41 comprises an oblong aperture 51 in which the end of the sheath 42 fixed by means of the nuts 43 is engaged. This oblong opening 51 makes it possible to adjust the alignment of the remote control relative to the rack 47 for controlling the pushing device. A bracket 53 fixed on one of the studs 38 and forming a support part 33 carries a fixing body 54 of a housing 55 in which are disposed the control pinions for the device for pushing the rods. The rack 47 engages, inside the housing 55, with a first pinion 52 mounted in the housing by means of a shaft on which is placed a second pinion 56 engaging with a third pinion 57 which is also mounted in the housing 55 by means of ball bearings 59. A tubular support 60, in which are fixed two ball bushes, is integrally mounted with the bracket 53 of the support 33 in a vertical direction. A shaft 61 is mounted so as to slide in the vertical direction inside the ball bush 60 and comprises a central part 61' forming a rack on which the pinion 57 engages. The actual pushing element of the device for pushing the rods consists of two support forks 62, 62' made integral by means of screw 63 of the body 54 fixed on the bracket 53 and a pushing fork 64 disposed between the support forks 62 and 62' and made integral by screws 65 of the sliding shaft 61. The support forks 62 and 62' each comprise an end part comprising two inner recesses 67, 67' which enable the support forks 62, 62' to engage around the guide tubes 3 in their end part located below the upper joining piece 4. In their operating position, as shown in FIG. 6, the forks 62 and 62' of the pushing element rest on the lower surface of the upper joining piece 4 of the assembly. The pushing fork 64 comprises, at its end, an inner recess 66 which can be engaged on the small-diameter part of a plug forming the end part of a peripheral rod 6a on which pushing is effected by means of the device shown in FIGS. 6, 7 and 8. By pushing on the flexible element 45 of the remote control, the rack 47 is displaced, driving the pinion 52 and, by means thereof, the pinions 56 and 57, in rotation. The pinion 57 drives the shaft 61 in a downward-directed vertical displacement which is transmitted to the pushing fork 64. The end of the pushing fork 64, engaging with the shoulder of the plug of the peripheral rod 6a, transmits a downward-directed pushing action to this rod. The device is positioned in its operating position, shown in FIG. 6, by the crossed carriage assembly 32 under the control of video cameras, such as 34, providing the operator placed in the control station 25 with an image of the zone in the vicinity of the upper end of the rod 6a. FIG. 6A shows the position of a pushing fork 64' which makes it possible to exert an upward-directed pushing action on the lower end of a peripheral rod 6a in the vicinity of the lower joining piece 5 of the assembly. The fork 64' has an identical form to the fork 64. This fork is mounted, in a position which is reversed relative to the position of the fork 64, on the body of the pushing device. The support forks 62, 62' rest on the upper surface of the lower joining piece 5 of the assembly. FIG. 9 shows an assembly for the control and measurement of displacement and pushing force which is denoted generally by the reference 70, this control and measurement assembly being fixed on the railing 37 of the platform 26 of the control station 25. This assembly 70 comprises a casing 71 of tubular form which is closed at its ends by end plates 72 and 73. The cylindrical housing 71 comprises, in its upper part shown on a larger scale in FIG. 9A, oblong openings 74 and 75 permitting the longitudinal displacement of corresponding indicators 76 and 77 as well as two measurement scales 78 and 79 comprising graduations which permit direct referencing of the displacements of the indicators 76 and 77, respectively. The graduated rule 78 makes it possible to evaluate, in the manner which will be described hereinbelow, the displacement of the rod as a function of the pushing force exerted on its end. The graduated rule 78 is fixed on the cylindrical body 71 by means of screws 80 engaged in oblong openings in the graduated rule 78. In its position shown in FIG. 9A, the graduated rule 78 makes it possible to measure the displacement of a pencil under the effect of a pushing action on its lower part, resting on the lower joining piece 5. The lefthand end of the rule 78 coincides with a reference I marked on the upper part of the cylindrical body 71. The rule 78 is graduated in millimeters, from 0 to 50 mm. In order to measure the displacement of a rod by pushing at the level of the upper joining piece, it suffices to turn the graduated rule 78 over by removing the screws 80, the righthand end of the graduated rule 78 coinciding with the reference S marked on the body 71. The oblong openings make it possible to perform a certain adjustment of the position selected as origin, as will be described hereinbelow. A second graduated rule 79 is fixed by screws 83 engaged in oblong openings in the rule 82 inside the body 71. The rule 82 comprises two graduated symmetrical parts which make it possible to measure the pushing force exerted on the rod by virtue of the position of the indicator 77. One of the scales corresponds to a pushing action on a fuel rod at the level of the lower joining piece and the other graduated scale corresponds to the pushing action on a rod at the level of the upper joining piece. A position of origin 84 is indicated on the body 71. As may be seen in FIG. 9, a profiled end 90 of a control screw is mounted so as to project outwards relative to the end plate 72. A second profiled end 91 of a control screw, shown in dotted lines, is disposed inside the casing 71, this end being accessible via an opening 93 in the end plate 72, which may be seen in FIG. 10. The casing 71 of the control and measurement assembly 70 is fixed on the railing 37 by means of a gripping yoke 94 comprising an upper flange 95 fixed on the casing 71 and a lower flange 96 which can be clamped against the flange 95 by screws 97 after engagement of the two half flanges on the railing 37. A joining cable 98 makes it possible to connect the two half flanges when the screws 97 are removed. A plate 99 then retains the screws 97 which thus cannot be lost. The plate 99 comprises openings for the passage of a wrench for tightening the screws 97. As may be seen in FIGS. 10, l0A, 11 and 12, two guide columns 101 and 101, are mounted in the longitudinal direction of the casing 71. An inner movable assembly 100 is mounted so as to slide by means of ball bushes 103 and 103' on the columns 101 and 101'. The movable assembly 100 comprises two end plates 104 and 105 on which are fixed the ball bushes, such as 103 and 103'. A maneuvering screw 106, integrally attached at its end to the maneuvering part 91, is mounted so as to rotate, by ball bearings, in the flanges 104 and 105. The screw 106 is mounted in the longitudinal direction in a position which is offset relative to the axis of the casing 71, located perpendicular to the guide column 101. A nut 107 comprising an internal thread is engaged on the screw 106 so that the nut 107 can be displaced by rotating the screw 106 from the control profile 91. The nut 107 is engaged in an opening of corresponding form provided between two parts which are assembled together of a lateral carriage 108 mounted so as to slide, by virtue of ball bushes 109, on the column 101. The position measurement indicator 76 is fixed on the lateral carriage 108 so that its end penetrates into the longitudinal aperture 75 of the casing 71. The end of the sheath 42 of the ball remote control opposite to the end connected to the pushing device, shown in FIGS. 6, 7 and 8, is fixed on the end plate 73 by means of a threaded part 110 on which a fixing nut 110' is engaged. The end of the sheath and the flexible element 45 pass through the flange 105 of the inner movable assembly 100 at the level of an opening 111. The end of the flexible element 45 projecting relative to the sheath 42 beyond its part 109 is fixed, by virtue of a threaded part and a nut 113, to the lateral carriage 108. The carriage 108 may be displaced in the longitudinal direction of the casing 71 by screw 106 which may be rotated by inserting a tool in the opening 93 of the end plate 72 of the casing, the tool being engaged on the control profile 91. The carriage 108 is guided in its displacements by the ball bushes 109 mounted so as to slide on the column 101. The displacements of the carriage 108 result in a displacement of the flexible element 45 inside the sheath 42, either in the direction of the pushing action or in the direction of the pulling action, according to the direction of rotation imparted to the screw 106. The indicator 76 makes it possible to reference the displacements of the carriage 108 and thus the displacements of the flexible element 45 of the ball remote control inside the sheath 42. Reference will now be made to FIGS. 11 and 12 in order to describe the part of the device which makes it possible to displace the inner movable assembly 100 inside the casing 71. A threaded shaft 115 is mounted so as to move in rotation, along the entire length of the casing by virtue of rolling bearings 116 and 117 fixed in the end plates 72 and 73. An end of the threaded shaft 115 projecting outwards relative to the end plate 72 is integrally attached to the control profile 90 on which it is possible to engage a tool in order to rotate the threaded shaft in one direction or another. A nut 118 comprising an internal thread, on which the screw thread of the shaft 115 is engaged, is placed in a central position between two support dishes 119 and 120. The movable assembly 100 comprises a sleeve 121 fixed, at its ends, to the flanges 104 and 105, respectively. Support plates 124 and 125 are fixed by screws to the outer face of the flanges 104 and 105, respectively. Two helical springs 127 and 128 are inserted between a support dish 119 (or 120) and the corresponding closure plate 124 (or 125). A rod 129, at the end of which the indicator 77 is fixed, passes through the wall of the sleeve at the level of an aperture 130 of longitudinal direction. Stops 131 and 132 made from a flexible material are fixed on the outer faces of the plates 124 and 125, respectively, in order to dampen the impacts on the end plates 72 (or 73) in the event of a faulty maneuver during displacement of the movable assembly 100. Rotation of the threaded shaft 115 in one direction and in the other, by using a tool engaged on the control profile 90, produces a displacement of the nut 118 and of the support dishes 119 and 120 so as to compress one of the two springs 127 and 128. The pushing action on the corresponding flange 104 (or 105), by means of the support plate 124 (or 125), produces a displacement of the movable assembly 100 and thus of the lateral carriage 108 when the compression of the spring produces a force which is sufficient to displace the carriage 108 and the movable flexible element 45, exerting a pushing action, by means of the pushing fork 64, on a peripheral rod 6a. The displacement of the indicator 77 opposite the graduated rule 79 makes it possible to determine the pushing force, the spring 127 (or 128) which is compressed during displacement of the nut 118, acting as a dynamometric spring. FIGS. 13 and 14 are charts which make it possible to determine the displacement of the pushing fork 64 as a function of the displacement of the indicator 76 and of the pushing force measured by virtue of the indicator 77 moving opposite the graduated rule 79. In fact, the values of the displacements read on the graduated rule 78 do not correspond to the actual value of the displacement of the fuel rod 6a, bearing in mind the adjustment of the mechanical play of the device and due to the fact that the ball remote control is activated by pushing on the flexible element 45. For example, in the case of FIG. 13 which relates to the pushing action on a fuel rod at the level of the upper joining piece of the assembly, a displacement of 25 mm of the indicator 76 on the graduated rule 78 corresponds to an actual displacement of the pushing fork 64 and of the rod 6a of 9.5 mm for a force of 50 daN. FIG. 14 relates to the displacement of a fuel pencil which is pushed by means of a pushing fork 64 at the level of the lower joining piece of the assembly. In the case of FIG. 14 which relates to a pushing action at the level of the lower joining piece of the assembly, a displacement of 25 mm of the indicator 76 on the graduated rule 78 corresponds to a displacement of the pushing fork 6 and of the fuel rod 6a of 8 mm for a force of 50 daN. A description will now be given, with reference to all the drawings figure, of a pushing operation on a peripheral rod 6a which makes it possible to measure precisely the axial retention force on this rod in the assembly. Before commencing the operation, the pushing device, for example such as shown in FIG. 6, is equipped with support forks 62, 62' and with a pushing fork 64 so that it is capable of pushing the peripheral fuel rods 6a either from the upper joining piece or from the lower joining piece. It is thus appropriate to select a pushing fork 64 whose form may be different according to the types of plugs for the rods being pushed. It is also appropriate to place this fork and the corresponding support forks in the desired position on the pushing support 61 formed by the sliding shaft and on the support in which the ball bushes 60 are fixed, respectively. The pushing fork is fixed in the desired direction, inside a housing provided on the sliding shaft 61 in its central part comprising the rack 61'. The pushing fork is fixed by means of the two screws 65. Then the entire displacement and measurement device placed at the control station and shown, in particular, in FIGS. 9 and 9A, is adjusted. The graduated rule 78 is mounted which makes it possible to measure displacement, in the desired direction according to the type of operation chosen, i.e., in the case of a pushing action from the upper joining piece of the assembly or in the case of a pushing action from the lower joining piece. The position of the adjuster is defined according to the circumstances by the reference I or the reference S marked on the surface of the casing 71. The rule 79 comprising the graduation for measuring the force exerted (in daN) is then fixed. The graduation 0 of the corresponding scale (pushing action from the upper joining piece or pushing action from the lower joining piece) is placed opposite the reference 84. By activating the control 91 of the screw 106, the indicator, 76 is displaced over a length of approximately 1 centimeter in the direction indicated by the arrow on the graduated rule 78. This adjustment operation must be performed without load, no force being applied on the pushing fork 64. This adjustment operation must be effected while the pushing tool is located inside the pool in a zone where no mechanical member or other obstacle risks coming up against the pushing fork. A check is then made that the indicator 77 is located opposite the graduation 0 of the scale corresponding to the envisaged pushing operation. If there is an offset relative to 0, the fixing screws 83 are loosened and the rule 79 is displaced in the desired direction before the screws 83 are retightened. The pushing fork 64 is then accurately positioned, the pushing assembly being fixed on the support 33 which is itself placed on the displacement device 32 which is integral with the carriage 30. In this phase, use is made of the control 91 of the screw 106 disposed inside the casing 71 in order to displace the indicator 76 up to the time when it coincides perfectly with the 0 of the graduation of the displacement measurement rule 78. A check is made and, if appropriate, the indicator is reset to 0 on the corresponding dynamometric scale. The pushing fork is then brought close to and engaged in the plug of the pencil. The support forks are placed in contact with the face directed towards the fuel pencils of the joining piece from which the pushing action is effected. This operation is performed by using the displacement means 30 and 32 and the means for visualizing the work zone, such as the video camera 34. The pushing fork 64 is brought perpendicular to the plug of the rod 6a by using the control 91 located inside the casing which drives the screw 106 in rotation, permitting the displacement of the carriage 108 on which is fixed the end of the flexible element 45 which controls the pushing fork. The indicator 76 is displaced relative to its initial position indicating the 0 on the graduated rule 78. The notched end part 66 of the pushing fork 64 is engaged in the groove of the plug of the rod 6a by maneuvering the displacement device 30, 32. The position of the indicator 76 on the graduated rule 78 is noted. For example, if a work operation is to be performed at the level of a lower joining piece, the indicator 77 indicating the position 0 on the dynamometric rule 78, a displacement of 10 mm of the indicator 76 for a force of 0 daN corresponds to an actual path of the pushing fork equal to 6.5 mm, as appears in FIG. 14. It may be deduced therefrom that the fuel rod 6a is placed at 6.5 mm beyond its extreme lower position. The fuel rod is then pushed and the pushing forces for obtaining a displacement of the rod are measured. To this end, use is made of the control 90, disposed outside the casing 71, in order to rotate the threaded shaft 115 in the desired direction. A displacement of the nut 118 and of the support dish is thus produced, on the corresponding spring 127 (or 128). The spring is compressed between the washer and the corresponding support plate of the movable assembly 100 and produces a pushing force which corresponds to the pushing force exerted by the fork 64 on the rod 6a. When the force exerted on the dynamometric spring is sufficient to overcome the axial retention force on the pencil in the assembly, the rod is displaced over a certain height so that the movable assembly 100 is capable of being displaced, which ensures decompression of the dynamometric spring. The displacement of the movable assembly 100 and of the pushing fork 6 is arrested due to the fact that the force exerted by the spring decreases during displacement of the movable assembly 100. The operator 23, located at the control station 25, makes a certain number of readings, by directly reading the position of the indicators relative to the scales of the graduated rules fixed on the upper part of the casing 71. He notes the force indicated by the indicator 77 on the rule 79 at the time when the rod begins to be displaced, i.e., the time when the indicator 76 leaves the position 0 on the graduated rule 78. This measurement of force in daN indicates the maximum value of the retention force on the rod. When the inner movable assembly 100 has completed its displacement, the graduated rule 79, which is fixed on the sleeve 121 of this movable assembly has been displaced, so that the indicator 77 indicates the actual value of the axial retention force on the rod in daN. The operator notes the values indicated by the indicators 76 and 77 on the graduated rules 78, 79, respectively. The results obtained are then analyzed with the aid of charts, such as shown in FIGS. 13 and 14. For example, in the case of a work operation performed from the lower joining piece of the assembly, the following values were noted: the indicator 77 indicates the value 20 daN on the dynamometric rule 79. PA1 the indicator 76 indicates the value of 18 mm on the rule 78 graduated in millimeters. A displacement of the indicator 76 of 18 mm, under a load of 20 daN, corresponds to a displacement of the rod of 8 mm relative to its extreme lower position, as may be seen on the chart in FIG. 14. If reference is made to the example described hereinabove, in which the initial position of the pushing fork 64 indicated a displacement of 6.5 mm of the rod relative to its extreme lower position, it may be deduced therefrom that the rod has been displaced in reality by 8-6.5=1.5 mm, under a load of 20 daN. Overall results obtained make it possible to determine the state of the framework of the fuel assembly in respect of, in particular, the retention springs of the fuel rods in the cells of the spacer grids. The device thus makes it possible to ensure that the retention springs of the peripheral rods of the assembly correctly fulfill their function after several operating cycles. The device according to the invention makes it possible to check the axial retention force on the peripheral pencils of the assembly without having to remove the lower and upper joining pieces, this operation being performed under several meters of water in the storage pool for the fuel. Implementation of the device is relatively simple in that the assembly for displacement, adjustment and maneuvering of the device used may be positioned and controlled from the edge of the pool. After each measurement, the rods which have been pushed may be re-placed in their initial position by using the remote control in the pulling direction, after having replaced a pushing tool fitted at the level of the joining piece opposite to that from which the initial pushing was performed. For this purpose, use could be made of two different pushing tools each associated to a carriage of a double crossed carriage device, one of the tools ensuring the initial pushing and the other tool the re-placement of the pencil from the opposite joining piece. If the measurement results obtained reflect displacements of the rods of a significant length under a small load, it may be deduced therefrom that the springs are no longer fulfilling their function of axially retaining the rods. In this case, it is appropriate to refuel the fuel rods in a new framework enabling them to be correctly retained. It is possible to use axial pushing devices and remote control means for these pushing devices for the rods which are produced in a form other than that which has been described. The means for measurement of the axial pushing force on the pencil and of the length of displacement may be different from those which have been described and comprise, for example, displacement sensors and strain gauges. Use may also be made, if appropriate, of the device according to the invention in order to check any inner row of rods of the assembly after removing the rods located on the outside of this inner row. It is then simply necessary to use a pushing fork of a length which is adapted in order to reach the rods which are at a distance from the periphery and even the rods located in the central part of the assembly from the time when these rods are accessible between the guide tubes of the assembly.
050842366	abstract	A nuclear reactor coolant pump for pumping reactor coolant fluid in a reactor coolant system includes a casing, a rotor and an impeller. The casing defines an inlet nozzle for receiving the fluid, a peripheral outlet nozzle for discharging the fluid, and a passage interconnecting the inlet and outlet nozzles through which the fluid can flow from the inlet to outlet nozzle. The rotor extends axially through the casing and has an end disposed adjacent passage. The impeller is mounted to the end of the rotor and disposed in communication with the passage and offset axially from the outlet nozzle of the casing. The impeller is rotatable with the rotor for drawing fluid into the casing through the inlet nozzle and discharging fluid from the casing tangentially through the peripheral outlet nozzle after flow through the passage. The outlet nozzle is composed of first and second wall portions defined above and below a plane extending generally parallel to the rotation axis of the impeller. The first wall portion has a substantially semi-cylindrical shape, whereas the second wall portion has a combined substantially semi-elliptical and semi-conical shape.
	claims	1. A system for capturing a radiation image of an object, the system comprising:a wavelength converter configured to generate scintillation light in response to incidence of radiation transmitted through the object;a first imaging device including an image sensor and configured to capture an image of scintillation light emitted from an incidence surface of the wavelength converter; anda stage configured to adjust a position of the first imaging device, wherein the stage is configured to rotate the first imaging device. 2. The system according to claim 1, wherein the stage is configured to rotate the first imaging device so as to maintain an angle formed between an optical axis of the first imaging device and the incidence surface of the wavelength converter. 3. The system according to claim 2, wherein the angle is 90°. 4. The system according to claim 1, wherein the stage has a rotation axis on the incidence surface of the wavelength converter. 5. The system according to claim 4, further comprising a radiation source configured to emit radiation toward the object, wherein the rotation axis intersects with an axis of the radiation source. 6. The system according to claim 4, wherein the rotation axis intersects with an optical axis of the first imaging device. 7. The system according to claim 1, further comprising a second imaging device including a second image sensor and configured to capture an image of scintillation light emitted from a surface opposite to the incidence surface of the wavelength converter. 8. The system according to claim 1, wherein the radiation is X-ray. 9. The system according to claim 1, wherein the wavelength converter comprises a scintillator. 10. A method for capturing a radiation image of an object, the method comprising:adjusting a position of a first imaging device including an image sensor,irradiating the object with radiation emitted from a radiation source;converting the radiation transmitted through the object to scintillation light by a wavelength converter; andcapturing an image of the scintillation light emitted from an incidence surface of the wavelength converter by the first imaging device,wherein the adjusting rotates the first imaging device. 11. The method according to claim 10, wherein the adjusting rotates the first imaging device so as to keep an angle based on an optical axis of the first imaging device and the incidence surface of the wavelength converter. 12. The method according to claim 11, wherein the angle is 90°. 13. The method according to claim 10, wherein the adjusting rotates the first imaging device on a rotation axis on the incidence surface of the wavelength converter. 14. The method according to claim 13, wherein the rotation axis intersects with an axis of the radiation source. 15. The method according to claim 13, wherein the rotation axis intersects with an optical axis of the first imaging device. 16. The method according to claim 10, further comprising capturing an image of scintillation light emitted from a surface opposite to the incidence surface of the wavelength converter. 17. The method according to claim 10, wherein the radiation is X-ray. 18. The method according to claim 10, wherein the converting is performed with the wavelength converter comprising a scintillator.
062787667	summary	BACKGROUND AND SUMMARY OF THE INVENTION The use of heavy metal collimators of circular shape is now well known for stereotactic radiosurgery using treatment planning machines such as linear accelerators (LINACs) as an X-ray source (see the XKnife information from Radionics, Inc., Burlington, Mass.). Circular collimators are used made of lead or Cerrobend heavy metal with circular apertures of different sizes to collimate the X-ray beams from a LINAC. A collimator is rotated in a so-called gantry angle and couch angle around an isocenter at which position is located a target volume within the body of a patient. Conformal stereotactic radiosurgery involves use of irregularly shaped collimators that are typically non-circular. These may be so-called cut-block collimators, multi-leaf collimators, or miniature multi-leaf collimators (see the information from Radionics, Inc., Burlington, Massachusetts or Fischer GmbMH, Frieburg, Germany). Conformal collimators are usually used in a static mode, meaning static discrete beam directions are determined and different collimators shapes are used depending on the shape of the target volume such as a tumor in the patient's head. Circular collimators are usually used in an arc mode, which means that the circular collimator is swept over the patient's head through the couch and gantry angles. A certain degree of target volume dose shaping is achieved by circular collimator arc therapy, but this is limited because of the limitation in shapes of the circular collimators. More conformal collimation is achieved by the cut-block or multi-leaf changeable shape collimators, but these are complicated devices and are labor intensive to make for a specific patient. In general, the system of the present invention is directed at an improved system for accomplishing conformal arc therapy for LINAC radiosurgery in the body. The system offers a simple and practical way of improving the dose distribution of X-rays for an irregularly shaped target volume by a combination of circular collimators and collimator blocking jaws which can be used to eclipse a portion of the circular beam aperture of the circular collimator. Heavy metal blocking jaws are typically used in the heads of the linear accelerator to provide large field blocking for standard radiotherapy irradiation of X-rays. Typically, a set of two pairs of opposing jaws orthogonally oriented to each other and moveable in an orthogonal direction to the beam direction are present in the gantry head of a typical X-ray LINAC. These jaws alone are normally not adequate to perform stereotactic radiosurgery. The penumbra effects of use of the four jaws in a LINAC combined with arc therapy would not provide sufficient tightness of radiation for small to medium size brain tumors for instance to be effective for radiosurgery and are typically not employed for such application in radiosurgery. Use of the straight jaw and circular collimator configuration are disclosed herein together with treatment planning software to accommodate its use for conformal arc radiosurgery.
	claims	1. A predictive model construction method of a reactor water radioactivity concentration predictive model construction device that predicts a radioactive metal corrosion product concentration in reactor water of a nuclear reactor in a nuclear power plant, the predictive model construction method comprising:the predictive model construction device executing:a step of calculating a plant state quantity prediction value to be calculated by using a physical model that describes plant state quantities of the nuclear power plant including a flow rate of feedwater and a metal corrosion product concentration in feedwater of the reactor water in the nuclear reactor; anda step of causing a machine learning model to learn learning data for supervised learning so as to construct a predictive model, the learning data for supervised learning including the plant state quantity prediction value and a plant state quantity that is able to be actually measured and includes at least one of the flow rate of feedwater, the metal corrosion product concentration in feedwater, a metal corrosion product concentration in reactor water, a radioactive metal corrosion product concentration in reactor water, a flow rate of reactor water cleanup system, duration of stay in reactor core in a fuel assembly, and electrical output of the nuclear reactor as input data, and including a radioactive metal corrosion product concentration in the reactor water as output data, wherein the radioactive metal corrosion product is an actual measured value, anda step of using the predictive model to predict radioactive corrosion product concentration on nuclear power plant structures and lower an exposure dose of work personnel during periodic inspection of the structures, whereinthe physical model is a mass balance model that uses the flow rate of feedwater and the metal corrosion product concentration in feedwater of the reactor water of the nuclear reactor and is related to a migration behavior of a metal corrosion product and a radioactive metal corrosion product in the reactor water, andthe plant state quantity prediction value includes at least one of an amount of metal corrosion product and an amount of radioactive metal corrosion product accumulated on a fuel rod of the nuclear reactor, the amount of metal corrosion product and the amount of radioactive metal corrosion product being calculated by using the mass balance model. 2. The predictive model construction method according to claim 1, whereinthe plant state quantity prediction value further includes at least one of a metal corrosion product concentration and a radioactive metal corrosion product concentration in the reactor water of the nuclear reactor, the metal corrosion product concentration and the radioactive metal corrosion product concentration being calculated by using the mass balance model. 3. The predictive model construction method according to claim 1, whereinthe radioactive metal corrosion product concentration in reactor water is a radioactive metal corrosion product concentration in reactor water of any of cobalt 60, cobalt 58, and manganese 54. 4. A prediction method of a reactor water radioactivity concentration prediction device that predicts a radioactive metal corrosion product concentration in reactor water of a nuclear reactor in a nuclear power plant,the prediction device stores a predictive model that is obtained by causing a machine learning model to learn learning data for supervised learning, the learning data for supervised learning including a plant state quantity that is able to be actually measured and includes at least one of a flow rate of feedwater, a metal corrosion product concentration in feedwater, a metal corrosion product concentration in reactor water, a radioactive metal corrosion product concentration in reactor water, a flow rate of reactor water cleanup system, duration of stay in reactor core in a fuel assembly, and electrical output of the nuclear reactor and a plant state quantity prediction value calculated by using a physical model that describes plant state quantities of the nuclear power plant including the flow rate of feedwater and the metal corrosion product concentration in feedwater of the reactor water in the nuclear reactor as input data and including a radioactive metal corrosion product concentration in the reactor water as output data, wherein the radioactive metal corrosion product concentration is an actual measured value, the prediction method comprising:the prediction device executing:a step of calculating the plant state quantity prediction value based on the plant state quantities by using the physical model;a step of inputting the calculated plant state quantity prediction value and a plant state quantity that includes at least one of the flow rate of feedwater, the metal corrosion product concentration in feedwater, the metal corrosion product concentration in reactor water, the radioactive metal corrosion product concentration in reactor water, the flow rate of reactor water cleanup system, the duration of stay in reactor core in the fuel assembly, and the electrical output of the nuclear reactor into the predictive model as input data and calculating a radioactive metal corrosion product concentration in the reactor water as output data; anda step of using the predictive model to predict radioactive corrosion product concentration on nuclear power plant structures and lower an exposure dose of work personnel during periodic inspection of the structures, whereinthe plant state quantity to be input to the predictive model is actually measured data at an input time point, andthe plant state quantity to be input to the predictive model further includes a planned value from the input time point to a predicted time point of the radioactive metal corrosion product concentration in reactor water. 5. The prediction method according to claim 4, whereinthe prediction device stores a plan pattern of a planned value of a plant state quantity from the input time point to the predicted time point of the radioactive metal corrosion product concentration in reactor water, andthe plant state quantity to be input to the predictive model further includes a planned value of the plan pattern from the input time point to the predicted time point of the radioactive metal corrosion product concentration in reactor water. 6. The prediction method according to claim 1, whereinthe radioactive metal corrosion product concentration in reactor water is a radioactive metal corrosion product concentration in reactor water of any of cobalt 60, cobalt 58, and manganese 54. 7. A prediction method of a reactor water radioactivity concentration prediction device that predicts a radioactive metal corrosion product concentration in reactor water of a nuclear reactor in a nuclear power plant,the prediction device stores a predictive model that is obtained by causing a machine learning model to learn learning data for supervised learning, the learning data for supervised learning including a plant state quantity that is able to be actually measured and includes at least one of a flow rate of feedwater, a metal corrosion product concentration in feedwater, a metal corrosion product concentration in reactor water, a radioactive metal corrosion product concentration in reactor water, a flow rate of reactor water cleanup system, duration of stay in reactor core in a fuel assembly, and electrical output of the nuclear reactor and a plant state quantity prediction value calculated by using a physical model that describes plant state quantities of the nuclear power plant including the flow rate of feedwater and the metal corrosion product concentration in feedwater of the reactor water in the nuclear reactor as input data and including a radioactive metal corrosion product concentration in the reactor water as output data, wherein the radioactive metal corrosion product concentration is an actual measured value, the prediction method comprising:the prediction device executing:a step of calculating the plant state quantity prediction value based on the plant state quantities by using the physical model;a step of inputting the calculated plant state quantity prediction value and a plant state quantity that includes at least one of the flow rate of feedwater, the metal corrosion product concentration in feedwater, the metal corrosion product concentration in reactor water, the radioactive metal corrosion product concentration in reactor water, the flow rate of reactor water cleanup system, the duration of stay in reactor core in the fuel assembly, and the electrical output of the nuclear reactor into the predictive model as input data and calculating a radioactive metal corrosion product concentration in the reactor water as output data; anda step of using the predictive model to predict radioactive corrosion product concentration on nuclear power plant structures and lower an exposure dose of work personnel during periodic inspection of the structures, whereinthe physical model is a mass balance model that uses the flow rate of feedwater and the metal corrosion product concentration in feedwater of the reactor water of the nuclear reactor and is related to a migration behavior of a metal corrosion product and a radioactive metal corrosion product in the reactor water, andthe plant state quantity prediction value includes at least one of an amount of metal corrosion product and an amount of radioactive metal corrosion product accumulated on a fuel rod of the nuclear reactor, the amount of metal corrosion product and the amount of radioactive metal corrosion product being calculated by using the mass balance model. 8. The prediction method according to claim 7, whereinthe plant state quantity prediction value further includes at least one of a metal corrosion product concentration and a radioactive metal corrosion product concentration in the reactor water of the nuclear reactor, the metal corrosion product concentration and the radioactive metal corrosion product concentration being calculated by using the mass balance model.
	description	This invention relates in general to nuclear reactors and, in particular, to sensor inserts employed in one or more internal components of a nuclear reactor, which house one or more environmental sensors. The fission reactions in a nuclear reactor generate heat and release neutrons which produce additional fission reactions in the nuclear fuel. The fissile material is massed in the reactor such that the neutron flux density is sufficient to maintain a sustained fission process. In a commercial reactor, pellets of the fissile material are encased in zircaloy rods mounted in modular, elongated fuel assemblies which are generally square in cross section. A large number of these square elongated fuel assemblies are arranged to form a generally cylindrical reactor core having a stepped periphery, which is housed inside a cylindrical stainless steel core barrel between horizontal upper and lower stainless steel core plates. This entire assembly is mounted inside a pressure vessel with generally hemispherical upper and lower heads. Reactor coolant, introduced into the pressure vessel through inlet nozzles, flows downward in an annular space between the core barrel and the pressure vessel, reverses direction in the lower plenum of the vessel, flows upward through openings in the lower core plate, and through the fuel assemblies where it is heated as a result of the fission reactions before being directed radially out of the pressure vessel through outlet nozzles. The heat extracted by the reactor coolant from the core is utilized to generate electricity thereby lowering the temperature of the reactor coolant which is recirculated through the reactor in a closed loop. Since the fuel assemblies are square in cross section, an irregular space exists between the periphery of the core and the inner surface of the core barrel. The usual practice is to place longitudinally extending flat, baffle plates along the outer surfaces of the fuel assemblies to confine the upward coolant flow to the fuel assemblies. The baffle plates are held in place by horizontal, irregularly shaped former plates that are bolted to and between the longitudinal baffle plates and the core barrel. Holes in the former plates permit limited coolant flow in the generally annular space between the longitudinal baffle plates and the core barrel to provide cooling for these components and to equalize the pressure on both sides of the longitudinal baffle plates. With an aging fleet of reactors around the world, there is a current need to extend the life of these reactors. To obtain operating license extensions a reactor operator has to show that the reactor vessel and its internal components can safely withstand the harsh environment that an operating reactor experiences over the period of the license extension. Computational models are often constructed of the components within and around the reactor pressure vessel in order to calculate the radiation environment that those components experience during plant operation. The results of those calculations are then often used with material behavioral correlations in order to predict the post irradiation behavior of various metals under different pressure and temperature conditions. Traditionally, the reactor pressure vessel has been the primary component of interest in these types of analysis. To that end, surveillance capsules are standard components within the nuclear reactor. The surveillance capsules contain material samples, dosimetry, and maximum temperature monitors, which are periodically tested to validate the computational models. Ex-vessel neutron dosimetry is another system that is used to validate those computational models. The thermal behavior of the reactor vessel internals components, such as the baffle and former plates, which are subject to gamma ray interactions that lead to heat generation, are also of interest and can be calculated using similar computational models as those described previously. However, there is no device currently in existence that can be used to validate those calculations either for the thermal or radiation behavior of these components that is in close proximity to the baffle and former plates and other reactor vessel internal components such as the upper and lower core plates. Accordingly, it is an object of this invention to provide an in-situ sensor that can be employed to determine the environment and the effects of the environment on the reactor internal components to better understand the state of these components and to validate the computational models. These and other objects are achieved by a reactor internals component having a sensor insert supported within the reactor internals. The sensor insert has a head and an elongated shank extending from the head to a distal end. The shank has a hollow compartment that extends at least partially between the distal end and the head. The cross sectional profile of the elongated shank is sized to fit into an opening in one or more reactor internals components. One or more self-contained, passive environmental sensors are secured within the hollow compartment which is closed off at the distal end by an end plug which is affixed to the shank. In one embodiment, the environmental sensors comprise a plurality of environmental sensors, respectively configured to monitor different environmental parameters. Preferably, the one or more environmental sensors comprise material samples, dosimetry or maximum temperature monitors. Also, it is desirable that the reactor internals sensor insert includes one or more coded markings that can be used to identify the location of the sensor insert within the reactor internals and, preferably, also identify the orientation of the sensor insert within the reactor internals. In another embodiment, the reactor internals sensor insert includes an anchor for fixing the elongated shank in the opening in the one or more reactor internals components. Preferably, the anchor comprises one of either a male or female thread extending over at least a portion of the shank, which is sized to mate with another of a male or female thread on the opening in the one or more reactor components into which the sensor insert is to be seated. Desirably, the anchor includes a locking mechanism that fixes an orientation of the shank within the opening in the one or more reactor internals components. The locking mechanism, for example, may be a lockbar that extends through an opening in the head and through a groove on the surface of the shank and partially into a groove in a surface of a wall on the opening in the one or more reactor internals components. Desirably, the lockbar is held in position within the opening in the head by a spring clip wedged against a portion of the head. The spring clip, for example, may be wedged in a counter bore in a surface of the head. In such case, desirably, the spring clip is a circular spring clip. In still another embodiment, the shank has an axial dimension along an elongated dimension of the shank and the hollow compartment is partitioned into separate axial compartments in which the environmental sensors are respectively supported. Similarly, the hollow compartment may be partitioned into separate circumferential compartments in which the environmental sensors are respectively supported. The one or more environmental sensors may also be housed within a partitioned sheath within which the sensors are separated, with the sheath sized to slide into and out of the hollow compartment. The sheath preferably has a positioning feature that fixes the orientation of the sheath relative to the hollow compartment and desirably the sheath has a coded marking that can be used to identify the reactor internals sensor insert in which it resided. The sheath may also include a gripping feature proximate the distal end to ease movement of the sheath out of the hollow compartment. The one or more environmental sensors may include a neutron activation wire that is enclosed within a stainless steel tubing, preferably with cadmium shielding. The shank preferably comprises Stainless Steel 347 and/or titanium. Preferably, the distal end of the shank has a larger circumference than a portion of the shank that extends from the enlarged head. In another embodiment, the head has a flared circumferential extension configured to mechanically engage with a circumferential machine groove or slot in the opening in the one or more reactor internals components. Typically, reactor vessel internals components are bolted together as a means of attachment. However, the number of bolts and bolting patterns are often conservatively specified, meaning that some bolts are extraneous. As such, these bolts can be replaced by non-load-bearing bolts that are designed to provide an accurate characterization of the temperature and radiation environment in and around the reactor vessel internals components. One embodiment of the sensor insert of this invention, to replace one or more of such bolts, is the fastener 10, illustrated in FIG. 1, which is a bolt-type of a device that can be used to fasten together one or more reactor vessel internals components. As such, this embodiment has a partial or full length thread 12 along its shank 14 where the thread extends only partially along the shank 14. The thread 12 extends from a distal end 18 and terminates at a reduced diameter section 16. The opposite end of the reduced diameter section 16 is connected to the underside of the bolt head 20. The opposite side of the bolt head 20 is provided with a slot 22 to aid in tensioning the bolt 10. The embodiment illustrated in FIG. 1 can have either or both a dimple 74 orientation mark and/or an identification/serial number 24 engraved to allow visual confirmation of the location and orientation of the bolt. Furthermore, the dimple orientation mark and/or an identification/serial number can allow personnel preparing the device to accurately record where and with what the device was loaded and to allow personnel unloading the device for post irradiation analysis to accurately characterize each component contained within the device. Internal features may also have the dimple orientation mark and/or an identification/serial number to ease the burden of record keeping. Referring to FIG. 2, it can be appreciated that the shank 14 has a hollow cavity 28 extending from the distal end 18 towards the head 20 a distance that is dictated by the space required for a number of monitoring sensors that will be used to characterize the reactor environment within the vicinity in which the bolt 10 is placed. FIG. 2A is a schematic of the bolt shown in FIG. 1 with the end cap 26 that closes off the hollow cavity 28, shown withdrawn from the opening. FIG. 2B is an end view of FIG. 2A taken at the distal end 18. FIG. 2C is the end view shown in FIG. 2B with the end cap 26 removed, exposing a partitioned sensor module 30 enclosed within a metal sheath 32. FIG. 2D is a cross sectional view of FIG. 2A taken along the lines IID-IID thereof showing the sensors within the hollow cavity 28. The sensors are self-contained, passive devices that include one or more of a thermoluminescent dosimeter 34 or a LiF optical crystal; cadmium shielded, stainless steel clad, activation wire 36; eutectic maximum temperature melt wire 38; and unshielded, stainless steel clad, activation wire 40. A passive sensor is intended to refer to a sensory device that is dormant while in the reactor, i.e., not transmitting the information being acquired while within the reactor, out of the reactor, until removed from the reactor. FIG. 3 is a different perspective view of the fastener shown in FIG. 1 with the sensor module 30 partially withdrawn from the hollow cavity 28. As can be appreciated from FIG. 3, the sensors can be supported in circumferentially and axially divided compartments. The sensor module 30 can be keyed as shown at 42 to ride in an axial groove in the sidewall of the hollow cavity 28 to fix the orientation of the sensor module. The outer sheath 32, of the sensor module 30 shown in FIG. 3, has been partially removed to expose the sensors into view. Accordingly, the fastener sensor of this embodiment has one or more cavities that may contain one or more sensors such as: neutron activation wires surrounded by hypodermic stainless steel tubing; gamma ray dosimetry such as thermoluminescent detectors or LiF crystals; temperature monitors such as melt wires of various metals surrounded by a glass tube; reduced-scale material specimens, (e.g., Charpy or tensile); and/or one or more internal organizers to contain and position any or all of the foregoing. The device has enough internal space to provide adequate separation between any cadmium shielded neutron activation wires and unshielded neutron activation wires. Dimple marks can be used to determine alignment and rotation orientation relative to well-known reactor vessel internal features; provide location indexing to laboratory personnel loading or unloading device; and further differentiate the sensor bolt from traditional (unmarked) bolts. Preferably, all corners have fillets to reduce stresses during use and a stepped end plug is welded into the bolt to ensure a pressure seal. Desirably, the fastener is constructed from a material that is in whole or in part both resistant to corrosion and has an elemental composition that is conducive to dosimetry applications, e.g., Stainless Steel 347. The bolt 10 may also be constructed from a material that is resistant to corrosion and has an elemental composition that minimizes neutron activation, which will ultimately reduce shipping and handling burdens and the radiation dose to post-irradiation analysis personnel, e.g., titanium. An alternate embodiment is illustrated in FIGS. 4 and 5 and differentiates from the foregoing embodiments in that it employs a circumferentially flared feature 44 on the head 20 to mechanically engage with a circumferentially machined groove or other feature in a pre-existing vacant bolt hole. Like reference characters are used to identify corresponding components among the several figures. The embodiment shown in FIGS. 4 and 5 removes the need to perform field welding operations to install lockbars to secure the replacement fastener. The enlarged region 46 on the on the shank 14 of the fastener 10 that is shown as threaded is maintained with the same general profile to mate with the former plate female threads and is not turned down to the reduced diameter shank profile in order to maintain the shielding offered by that material to the shank in a traditional baffle bolt. In other respects, the embodiments illustrated in FIGS. 4 and 5 are similar to that previously described. FIGS. 6 and 7 illustrate another embodiment of the bolt-type fastener application of the sensor insert claimed hereafter, which is an alternate to that described previously with respect to FIGS. 1-5 and includes a mechanical locking feature 48, which in this case is a lockbar assembly, which mechanically engages and locks the orientation of the threads 12 once installed. This mechanical feature includes a lockbar that is circular, elliptical, or polygonal in cross section that is inserted through a hole 52 in the bolt head 20 and through a slot 54 bored axially through the thread 12 and through a corresponding slot bored through the female thread in the hole in the reactor internals component to which it is secured. The lockbar 50 is fixed in place with an outwardly expanding circular spring clip 58 that expands against a counter bore 56 in the bolt head 20. Alternately, the lockbar may be welded to the head to secure it in position. The lockbar assembly prevents rotation by engaging with and between the slot in the distal end 18 of the dosimetry bolt 10 and the slot through the female thread in the hole of the reactor internals component to which it is secured. The bolting arrangements in reactor internals components are conservatively specified; meaning that some bolts are extraneous. Preferably, the neutron dosimetry fastener embodiment of the sensor insert of this invention can replace non-load-bearing bolts, though the fasteners constructed in accordance herewith can carry some loads. With internal cavities 28 designed to hold neutron dosimetry, photon dosimetry, maximum temperature monitors and/or material specimens, sealed off, preferably with a welded plug, the contents are secured and protected from the harsh external environment. These cavities can be evacuated and back filled with an inert gas for further protection. In order to allow post-irradiation analysis, the plug 26 is welded to the distal end 18 and has a stepped axial contour that can be cut through and removed in a hot cell. In addition, the sensor insert, internal structures, and/or plug can be made of a material that is both resistant to corrosion and has an elemental composition that is conducive to dosimetry applications, e.g., Stainless Steel 347. Such a material would provide iron, nickel, cobalt, and niobium activation that could be counted later to characterize the neutron flux environment that the device resided within. Alternately, some or all of the sensor insert's material can be chosen to be both resistant to corrosion and have an elemental composition that minimizes neutron activation which will ultimately reduce shipping and handling burdens and the radiation dose to post-irradiation analysis personnel (e.g., titanium). The sensor inserts, constructed in accordance herewith, can be used in multiple different locations (radially, axially and azimuthally) to provide a three-D map of the radiation and/or temperature environment. These sensor inserts are particularly suited for use as baffle-former bolts, especially in a reactor vessel internals components that undergo significant irradiation and/or have experienced broken bolts thought to be caused by irradiation-induced material changes. The fasteners are designed to installed, affixed, and removed in an identical fashion to the bolts that they are replacing. However, they have identification information to allow plant operators to confirm positioning and orientation. There are also marks that allow laboratory personnel to load and unload the sensor inserts' contents while accurately recording the positions of the contents. Though the fastener embodiments were stated as particularly suited for use as baffle-former bolts, these types of dosimetry insert devices, i.e., installable devices encapsulating monitors for neutron, peak temperature, and/or gamma ray sensors that are inserted into a reactor internals component, can be used in place of a thimble plug to perform upper core plate monitoring. Such a device could also be inserted into an incore thimble tube and could be located to address any axial position in the lower internals. The foregoing are just examples of the many uses such a device can serve. For example, FIGS. 8 and 9 show one embodiment of a fuel assembly thimble plug assembly which can benefit from this invention. Thimble plug assemblies are inserted in unused thimbles in a fuel assembly located at a position in the core not accessed by the control rods, to restrict flow through some or all of those unused thimbles. As noted above, like reference characters are used to identify corresponding components among the several figures. FIG. 8 is a plan view of a thimble plug assembly 60 and FIG. 9 is an elevational view of the thimble plug assembly of FIG. 8 with one of the thimble plugs 62 (on the left side of the figure) replaced by a sensor insert whose sensor module 30 housed in its internal hollow cavity 28 (represented by the dotted lines) is identical to that described previously. Each of the thimble plugs 62 is supported by and suspended from a base plate 64. The base plate 64 has a vertically extending, centrally located sheath 66 that has an enlarged cap 72 that retains a slidably mounted bar assembly that is biased in an upward direction by a spring 68 wound around the sheath. When positioned in the core with the internals in place, the thimble plugs will be inserted into the empty fuel assembly guide tubes and the bar assembly 70 will be pressed against the upper core plate with the spring 68 compressed to prevent the thimble plug assembly from vibrating. Except for its outward configuration the sensor insert in the form of a thimble plug 62 is internally configured the same as the bolt-type embodiment 10. A major distinction between the surveillance capsules of the prior art and the sensor insert claimed hereafter is the positioning of the sensors within a reactor internals component or in place of a reactor internals component so that the sensors are intimately in proximity to the region of interest (e.g., within the baffle plates or the upper and lower core plates). In that way the sensors can register the actual environment that the material/component of interest is experiencing. In contrast, prior art surveillance capsules are positioned in order to irradiate the reactor vessel material samples at an accelerated rate relative to the reactor vessel itself in order to give advance analysis capabilities to determine how the reactor vessel material will behave later in life. However, the positioning also registers a relatively hard neutron spectrum because of the downcomer water further moderating the neutron flux at the reactor vessel relative to the surveillance capsule position on the outside of the core barrel. Also, any thermal gradient due to the water would not be reflected. Thus the surveillance capsules do not provide a truly one-to-one spectral and thermal comparison to what the reactor vessel is experiencing. For the purpose of monitoring the reactor vessel internals, which experience a yet harder spectrum, the sensor inserts of this invention are superior and eliminate any discrepancy in the spectral or thermal environment. While outwardly it may appear logical to position sensors for monitoring the internals as the prior art surveillance capsules are positioned, however, there are several reasons why that logic is not effective. In the first instance, in the prior art surveillance capsule position the irradiation rate of the sensors would lag that of the actual material/component of interest. Secondly the massive stainless steel reactor internals structures would to some extent shield the sensor from the neutron flux the actual material/component of interest was experiencing. Additionally, the temperature environment would be significantly different as there is a significant temperature differential over the radial extent of the reactor vessel internals. Furthermore, placing the prior art surveillance capsules on the outside of the core barrel provides a conservative reading of the environmental effects on the reactor vessel material. However, placing the reactor internals sensors on the outside of the core barrel, for the reasons noted above is non-conservative and would not provide information representative of what the internals are or will be experiencing. Furthermore, 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. For example, the sensors, rather than being loaded through the distal end, may be loaded through an opening in the head that is later sealed with a plug. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
	claims	1. A method of manufacturing a rotation detector that comprises a semiconductor chip, a nonmagnetic case body, a biasing permanent magnet and a nonmagnetic cap fixed to the case body for accommodating therein the semiconductor chip and the biasing permanent magnet, wherein: the case body has a joint surface; and the cap has an inside surface in contact with the joint surface, the method comprising:assembling the semiconductor chip, the case body and the biasing permanent magnet into a unit;forming at least one groove adjacent to the joint surface so as to form a space between the joint surface and the inside surface when the cap is fixed to the case body;fixing the cap to the case body so that the inside surface of the cap can be in contact with the joint surface;irradiating a laser beam at portions behind the inside surface of the cap under a prescribed pressure to melt materials of the case body and the cap and to make melted materials flow to the space;filling the melted materials in the space under the prescribed pressure;stopping irradiation of the laser beam when the melted materials function to gradually stop reducing a height of the inside surface of the cap from the joint surface of the case body; andcooling the melted materials to re-crystallize. 2. A method as in claim 1, wherein the laser transmittance of the case body is less than the laser transmittance of the cap. 3. A method as in claim 2, wherein:each of the case body and the cap is made of polyphenilene sulfide; andthe case body contains more carbon black than the cap. 4. A method as in claim 1, further comprising rotating the case body and the cap while irradiating a laser beam. 5. A method as in claim 4, wherein the pressure in irradiating is applied to the bottom end of the cap. 6. A method as in claim 1, wherein:the joint surface of the case body is tapered; andthe inside surface of the cap is shaped conical to be in close contact with the joint surface. 7. A method as in claim 1, wherein:the joint surface of the case body is tapered;the joint surface of the case body has a plurality of ring-shaped grooves, including the at least one groove, that circumferentially surround the joint surface on the opposite sides thereof; andthe inside surface of the cap is shaped conical and the conical inside surface confronts the tapered joint surface.
054886449	claims	1. A spring assembly for location between a pair of adjacent ferrules in a nuclear fuel bundle having a plurality of nuclear fuel rods passing through the ferrules and biasing the fuel rods of adjacent ferrules against stops in the ferrules, the spring assembly in an unstressed condition, comprising: first and second spring bodies lying in respective planes, each spring body having a central leaf, a pair of outer leaves spaced from said central leaf, and an end portion at each of opposite ends of said spring body joining ends of said central and outer leaves, said central leaf of each spring body having an intermediate portion projecting forwardly of the plane of said spring body to a fuel rod contacting forward side of said plane, said pair of outer leaves of each spring body having intermediate portions projecting rearwardly of the plane of said spring body, said spring bodies being disposed in back-to-back relation to one another with said intermediate portions of said outer leaves engaging one another, said planes lying parallel to and spaced from one another and said end portions lying in registration with one another whereby said intermediate portions of said central leaves project to opposite sides of said assembly for engagement with fuel rods in next-adjacent ferrules. a matrix of ferrules for receiving the fuel rods in said spacer, each ferrule having a plurality of fuel rod contacting points for abutting the fuel rods, adjacent pairs of said ferrules in said matrix comprising cylindrical members each having an axis, open opposite ends and a generally rectilinear opening in a side wall thereof, said adjacent pairs of said ferrules being disposed in side-by-side relation to one another with said rectilinear openings in lateral registration with one another; spring assemblies for location between said pairs of adjacent ferrules for biasing the fuel rods in said pair of ferrules into engagement with said contact points; each said spring assembly comprising first and second spring bodies lying in respective planes, each said spring body having a central leaf, a pair of outer leaves spaced from said central leaf, and an end portion at each of opposite ends of said spring body joining the ends of said central and outer leaves, said central leaf of each spring body having an intermediate portion projecting forwardly of the plane of said spring body to a fuel rod contacting forward side of said plane, said pair of outer leaves of each spring body having intermediate portions projecting rearwardly of the plane of said spring body, said spring bodies being disposed in back-to-back relation to one another, with said intermediate portions of said outer leaves engaging one another, said planes lying parallel to and spaced from one another, said end portions lying in registration with one another; said spring assemblies being disposed between said adjacent pairs of said ferrules with the intermediate portions of said central leaves lying in said rectilinear openings and said intermediate portions of said outer leaves lying between said adjacent pairs of said ferrules. a spring body having a central leaf lying in a plane, a pair of outer leaves spaced laterally from said central leaf, and an end portion at each of opposite ends of said spring body joining ends of said central and outer leaves, said central leaf of said spring body having an intermediate portion projecting forwardly of said plane to a fuel rod contacting forward side of said plane, said pair of outer leaves of said spring body lying rearwardly of the plane of said body and having intermediate portions projecting rearwardly of said outer leaves, said spring body being adapted for disposition between the adjacent ferrules with the intermediate portion of said central leaf projecting through an opening in the one ferrule for biasing the fuel rod therein against the stops and the intermediate portions of said outer leaves bearing against an outer surface of another ferrule of the adjacent ferrules. a matrix of ferrules for receiving the fuel rods in said spacer, each ferrule having a plurality of fuel rod contacting points for abutting the fuel rods, adjacent pairs of said ferrules in said matrix comprising cylindrical members each having an axis, open opposite ends and a generally rectilinear opening in a side wall thereof, said adjacent pairs of said ferrules being disposed in side-by-side relation to one another with said rectilinear opening in said one ferrule lying in lateral registration with a side wall of another ferrule of said adjacent pairs thereof; a spring body having a central leaf lying in a plane, a pair of outer leaves spaced laterally from said central leaf, and an end portion at each of opposite ends of said spring body joining ends of said central and outer leaves, said central leaf of said spring body having an intermediate portion projecting forwardly of the plane thereof to a fuel rod contacting forward side of said plane, said pair of outer leaves of said spring body lying rearwardly of the plane of said body and having intermediate portions projecting rearwardly of said outer leaves; said spring body being disposed between said adjacent ferrules, with the intermediate portion of said central leaf disposed in said rectilinear opening for biasing the fuel rod in the one ferrule against the stops and said intermediate portions of said outer leaves disposed between said adjacent ferrules and engaged against an outer surface of said another ferrule of said adjacent ferrules. 2. A spring assembly according to claim 1 wherein said end portions extend laterally beyond the outer leaves, terminating in end tabs. 3. A spring assembly according to claim 2 wherein said end tabs on each spring body project out of the plane of said body and to said forward side of said plane. 4. A spring assembly according to claim 1 wherein said intermediate portions of said outer leaves are welded in back-to-back relation to one another. 5. A spring assembly according to claim 1 wherein each said central leaf is resiliently flexible for bending movement in a rearward direction. 6. A spring assembly according to claim 1 wherein said spring bodies are formed of sheet metal. 7. A spring assembly according to claim 1 in combination with a pair of ferrules, each ferrule comprised of a cylindrical member having an axis, open opposite ends and a generally rectilinear opening in a side wall thereof, said ferrules being disposed in side-by-side relation to one another with said rectilinear openings in lateral registration with one another, said spring assembly being disposed between said ferrules with the intermediate portions of said central leaves disposed in said rectilinear openings and said intermediate portions of said outer leaves disposed between said ferrules. 8. The combination of claim 7 wherein said end portions extend laterally beyond the outer leaves and terminate in end tabs, said end tabs on each spring body projecting forwardly out of the plane of said body to overlie outer surfaces of said ferrules for laterally locating the spring assembly. 9. A spacer assembly for maintaining a matrix of nuclear fuel rods in spaced-apart relation between upper and lower tie plates, said spacer assembly comprising: 10. A spacer assembly according to claim 9 wherein said end portions of each said spring assembly extend laterally beyond the outer leaves terminating in end tabs, said end tabs on each said spring body projecting forwardly out of the plane of said body to overlie the outer surface of a ferrule of said adjacent pairs of ferrules. 11. A spring for location between a pair of adjacent ferrules in a nuclear fuel bundle having a plurality of nuclear fuel rods passing through the ferrules and biasing a fuel rod of one of the adjacent ferrules against stops in the one ferrule, the spring assembly in an unstressed condition, comprising: 12. A spring according to claim 11 in combination with the pair of adjacent ferrules, each ferrule comprised of a cylindrical member having an axis, open opposite ends and a generally rectilinear opening in a side wall thereof, said ferrules being disposed in side-by-side relation to one another with said rectilinear opening in said one ferrule in lateral registration with a side wall of another ferrule of said adjacent ferrules, said spring assembly being disposed between said adjacent ferrules with the intermediate portion of said central leaf disposed in said rectilinear opening for biasing the fuel rod in the one ferrule against the stops and said intermediate portions of said outer leaves disposed between said adjacent ferrules and engaged against an outer surface of said another ferrule of said adjacent ferrules. 13. A spacer assembly for maintaining a matrix of nuclear fuel rods in spaced-apart relation between upper and lower tie plates, said spacer assembly comprising: 14. A spacer according to claim 13 having a peripheral band about said ferrules within said spacer and openings in said band, pins received through said openings in said band and engageable with the central leaf and margins of the rectilinear opening in said one ferrule for compressing the spring and displacing the projection on said central leaf in a direction toward said plane of the spring body. 15. A spring according to claim 11 wherein said outer leaves lie in discrete planes angled relative to one another and intersecting at a location forwardly of the plane through said central leaf.
	claims	1. A system for achromatically bending a particle beam by about 90°, the system comprising:first, second, third, and fourth bending magnets serially arranged along a beam path of the particle beam,the first and fourth bending magnets being configured to generate a positive field gradient that defocuses the particle beam in a bend plane;the second and third bending magnets being configured to generate a negative field gradient that focuses the particle beam in the bend plane;the first, second, third, and fourth bending magnets collectively bending the particle beam by about 90°. 2. The system of claim 1, wherein the first, second, third, and fourth bending magnets each bend the particle beam by about 22.5°. 3. The system of claim 1, wherein the first and fourth bending magnets each comprise an iron cored, dipole electromagnet having pole faces that are inclined relative to one another and shaped so as to generate the positive field gradient, andwherein the second and third bending magnets each comprise an iron cored, dipole electromagnet having pole faces that are inclined relative to one another and shaped so as to generate the negative field gradient. 4. The system of claim 3, wherein the pole faces of the first, second, third, and fourth bending magnets are shaped such that the positive field gradient is substantially weaker than the negative field gradient. 5. The system of claim 3, wherein the pole faces of the first, second, third, and fourth bending magnets are each approximately hyperbolically shaped. 6. The system of claim 1, wherein the second bending magnet focuses the particle beam to a beam waist located reflection mirror plane midway between the second and third bending magnets. 7. The system of claim 1, further comprising an X-ray target configured to be irradiated with the bent particle beam. 8. The system of claim 1, wherein a midpoint between the second and third bending magnets defines a mirror plane, the first and fourth bending magnets being positioned substantially symmetrically across the mirror plane from one another, and the second and third bending magnets being positioned substantially symmetrically across the mirror plane from one another. 9. The system of claim 1, wherein the particle beam has a substantially round profile before entering the system and following bending. 10. The system of claim 1, wherein the particle beam comprises particles having an energy spread of 30% full width or less. 11. A method for achromatically bending a particle beam by about 90°, the system comprising:bending the particle beam with a first bending magnet that defocuses the particle beam in a first plane with a positive field gradient, and thenbending the particle beam with a second bending magnet that focuses the particle beam in the first plane with a negative field gradient, and thenbending the particle beam with a third bending magnet that focuses the particle beam in the first plane with a negative field gradient, and thenbending the particle beam with a fourth bending magnet that defocuses the particle beam in the first plane with a positive field gradient,wherein the first, second, third, and fourth bending magnets collectively bend the particle beam by about 90°. 12. The method of claim 11, wherein the first, second, third, and fourth bending magnets are serially arranged along a beam path of the particle beam, andwherein a midpoint between the second and third bending magnets defines a reflection plane, the first and fourth bending magnets being positioned substantially symmetrically across the reflection plane from one another, and the second and third bending magnets being positioned substantially symmetrically across the reflection plane from one another. 13. The method of claim 11, wherein the positive field gradient is substantially weaker than the negative field gradient. 14. The method of claim 11, wherein the first and fourth bending magnets each comprise an electromagnet and a core having a pole face that is inclined relative to a second plane and shaped so as to generate the positive field gradient, andwherein the second and third bending magnets each comprise an electromagnet and a core having a pole face that is declined relative to the second plane and shaped so as to generate the negative field gradient. 15. The method of claim 14, wherein the pole faces of the first, second, third, and fourth bending magnets are each approximately hyperbolically shaped. 16. The method of claim 11, wherein the first, second, third, and fourth bending magnets each bend the particle beam by about 22.5°. 17. The method of claim 11, wherein the second bending magnet focuses the particle beam to a beam waist located at the reflection plane. 18. The method of claim 11, wherein the particle beam has a substantially round profile before entering the system and has a substantially round profile following bending. 19. The method of claim 11, wherein the particle beam comprises particles having an energy spread of 30% full width or less. 20. A system for achromatically bending a particle beam by about 90°, the system comprising:first, second, third, and fourth bending magnets serially arranged along a beam path of the particle beam,a midpoint between the second and third bending magnets defining a reflection plane, the first and fourth bending magnets being positioned substantially symmetrically across the reflection plane from one another, and the second and third bending magnets being positioned substantially symmetrically across the reflection plane from one another;the first and fourth bending magnets being configured to generate a positive field gradient that defocuses the particle beam in a bend plane; andthe second and third bending magnets being configured to generate a negative field gradient that focuses the particle beam in the bend plane.
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	description	This application is a continuation of U.S. patent application Ser. No. 15/631,741, filed Jun. 23, 2017, now U.S. Pat. No. 10,497,479, which application claims the benefit of U.S. Provisional Patent Application No. 62/363,117, titled “VERTICALLY-SEGMENTED NUCLEAR REACTOR”, filed Jul. 15, 2016, which application is hereby incorporated by reference. The utilization of molten fuels in a nuclear reactor to produce power provides significant advantages as compared to solid fuels. For instance, molten fuel reactors generally provide higher power densities compared to solid fuel reactors, while at the same time having reduced fuel costs due to the relatively high cost of solid fuel fabrication. Molten fluoride fuel salts suitable for use in nuclear reactors have been developed using uranium tetrafluoride (UF4) mixed with other fluoride salts. Molten fluoride salt reactors have been operated at average temperatures between 600° C. and 860° C. Binary, ternary, and quaternary chloride fuel salts of uranium, as well as other fissionable elements, have been described in co-assigned U.S. patent application Ser. No. 14/981,512, titled MOLTEN NUCLEAR FUEL SALTS AND RELATED SYSTEMS AND METHODS, which application is hereby incorporated herein by reference. In addition to chloride fuel salts containing one or more of UCl4, UCl3F, UCl3, UCl2F2, and UClF3, the application further discloses fuel salts with modified amounts of 37Cl, bromide fuel salts such as UBr3 or UBr4, thorium chloride fuel salts, and methods and systems for using the fuel salts in a molten fuel reactor. Average operating temperatures of chloride salt reactors are anticipated between 300° C. and 800° C., but could be even higher, e.g., >1000° C. This disclosure describes various configurations and components of a molten fuel fast or thermal nuclear reactor. For the purposes of this application, embodiments of a molten fuel fast reactor that use a chloride fuel will be described. However, it will be understood that any type of fuel salt, now known or later developed, may be used and that the technologies described herein may be equally applicable regardless of the type of fuel used, such as, for example, salts having one or more of U, Pu, Th, or any other actinide. Note that the minimum and maximum operational temperatures of fuel within a reactor may vary depending on the fuel salt used in order to maintain the salt within the liquid phase throughout the reactor. Minimum temperatures may be as low as 300-350° C. and maximum temperatures may be as high as 1400° C. or higher. FIG. 1 illustrates, in a block diagram form, a simplified rendering of the basic configuration of a vertically-segmented nuclear molten fuel reactor. In general, a vertically-segmented molten fuel reactor 100 includes a reactor core vessel 104 defining a reactor core 102 of a fissionable fuel salt that is liquid at the operating temperature. Fissionable fuel salts include salts of any nuclide capable of undergoing fission when exposed to low-energy thermal neutrons or high-energy neutrons. Furthermore, for the purposes of this disclosure, fissionable material includes any fissile material, any fertile material or combination of fissile and fertile materials. The size of the reactor core vessel 104 may be selected based on the characteristics and type of the particular fuel salt being used in order to achieve and maintain the fuel in an ongoing state of criticality, during which the heat generated by the ongoing fission events generates heat in the fuel causing the temperature of the molten fuel to rise when it is in the reactor core vessel 104. In the vertically-segmented molten fuel reactor embodiments described in this application, a multi-stage heat exchanger is provided that includes at least one first heat exchanger stage 110 and at least one second heat exchanger stage 112 located above the reactor core vessel 104. The reactor core vessel 104, heat exchanger stages 110, 112, pump 114, molten fuel circulation equipment (including other ancillary components that are not shown such as piping, check valves, shutoff valves, flanges, drain tanks, ducts, flow directing baffles, etc.) and any other components through which the molten fuel salt circulates or contacts during operation can be referred to as the fuel circuit. In the fuel circuit, the hot fuel salt is circulated from the reactor core 102, up through the first heat exchanger stage 110, down through the second heat exchanger stage 112, and cooled fuel salt is returned into the core vessel 104. Fuel salt flow in the reactor core 102 is illustrated by the dashed arrows. For simplicity, only one first heat exchanger stage 110 and only one second heat exchanger stage 112 are illustrated in FIG. 1. Alternative embodiments of the vertically-segmented reactor, as will be discussed in greater detail below, may use any number and/or configuration of heat exchangers for either the first or second heat exchanger stage 110, 112. Furthermore, except were explicitly discussed otherwise, heat exchangers will be generally presented in this disclosure in terms of simple, single pass, shell-and-tube heat exchangers having a set of tubes and with tube sheets at either end. However, it will be understood that, in general, any design of heat exchanger may be used, although some designs may be more suitable than others. For example, in addition to shell-and-tube heat exchangers, plate, plate and shell, printed circuit, and plate fin heat exchangers may be suitable. In addition, although described in terms of having the fuel salt on the tube side and the coolant flowed through the shell side of the exchanger, this could be reversed in an alternative embodiment. For safety, all of the components of the fuel circuit are enclosed in a containment vessel 118. In an embodiment, the containment vessel 118 is a solid lower vessel portion 118A with no penetrations and a top portion 118B through which the reactor core and other components may be accessed. The lower vessel 118A completely surrounds the components of the fuel circuit such that a fuel salt leak from any component will be caught in the bottom of the containment vessel 118. In the embodiment shown, the circulation may be driven using one or more pumps, such as fuel salt pump 114. While fuel salt pumps 114 may be located anywhere in the fuel circuit, in the embodiment shown the pump 114 is located above the heat exchangers 110, 112 to pump fuel salt from the outlet of the first heat exchanger stage 110 to inlet of the second heat exchanger stage 112. In an alternative embodiment, the reactor core 104 and heat exchangers 110, 112 may be configured such that fuel circulation through the fuel circuit is driven by the density differential created by the temperature difference between the higher temperature fuel salt in the core 104 and the lower temperature salt elsewhere in the fuel circuit. This may be referred to as natural circulation. In many fuel salts, higher temperature molten salt is less dense than lower temperature salt. For example, in one fuel salt (71 mol % UCl4-17 mol % UCl3-12 mol % NaCl) for a 300° C. temperature rise (e.g., 627° C. to 927° C.), the fuel salt density was calculated to fall by 18%, from 3680 to 3010 kg/m3. Such a configuration obviates the need for fuel salt pumps 114. However, relying solely on natural circulation may limit the amount of heat that can be removed and thus limits the total power output of the reactor 100. The first and second heat exchanger stages 110, 112 transfer heat from the molten fuel salt to a primary coolant. In an embodiment the primary coolant may be another salt, such as NaCl—MgCl2. Other coolants are also possible including Na, NaK, supercritical CO2, lead, and lead bismuth eutectic. The primary coolant is circulated through a coolant circuit, such as by a pump 116. In an embodiment, the coolant may be maintained at a higher pressure so that any leakage in the fuel circuit will result in coolant entering the fuel circuit rather than fuel entering the coolant circuit. In the embodiment shown, primary coolant is circulated into the containment vessel 118 through the top portion 118B, through the first and second heat exchanger stages 110, 112, back out of the containment vessel 118, again, through the top portion 118B, and to a power generation system 120. The power generation system 120, as is known in the art, may be any type of system adapted to generate power from heated fluids. The performance of the reactor 100 may be improved by using one or more reflectors 108 to reflect neutrons back into the center of the core vessel 104 to assist in maintaining criticality within the reactor core section 102 and/or the breeding of fissile fuels from fertile feed materials. By reducing such losses of neutrons, the amount of fuel salt necessary for criticality, and therefore, the size of the reactor core 102, may be reduced. The reflector 108 may be formed from any material known in the art suitable for neutron reflection. For example, materials with reflective properties may include, but are not limited to, one or more of zirconium, steel, iron, graphite, beryllium, tungsten carbide, lead, lead-bismuth and like materials. The reflector 108 may be a single component or any number of separate elements containing some amount of reflective material. As the efficiency of the reflector 108 is affected by the amount of reflective material in the path of neutrons leaving the core 102, the reflector 108 may be of any design or shape as long as the desired amount of reflective material is provided. However, the efficiency is also affected by the amount of absorbing material, such as structural elements in reflectors 108 used to contain the reflective material, so certain design trade-offs need to be managed when designing and placing reflectors in a reactor 100. The reflector 108 may be outside of the core vessel 104 as shown, within the core vessel 104 (as further described below), or some combination of both. In the reactor embodiment shown, a reflector 108 separates the heat exchangers 110, 112 from the reactor core 104 with flow channels provided for the circulation of salt into and out of the core vessel 104. In a simple configuration, the reflector 108 may be a vessel containing a reflective material, such as lead, in which the reactor vessel 104 is located. In an alternative embodiment, the reflector 108 may include some number of reflector elements, such as tubes or containers filled with reflective material, spaced around the periphery of the reactor core vessel 104. It is noted that at some operating temperatures of the nuclear reactor 100 of the present disclosure a variety of neutron reflecting materials will liquefy. For example, lead and lead-bismuth are both materials that provide good neutron reflecting characteristics. However, lead melts at approximately 327° C., while lead-bismuth alloys commonly have melting temperatures below 200° C. As noted elsewhere in this application, the reactor 100 may operate in a temperature range from approximately 330 to 800° C., above the melting points associated with lead and lead-bismuth alloys. The reflector 108 or separate reflector elements may be formed from any material known in the art and may be selected based on consideration of any one or more design functions including temperature resistance, corrosion resistance, non-reactivity with other components and/or the fuel, radiation resistance, required structural support, weight, etc. In some cases, one or more reflector elements may be formed out of a structural material that holds or contains (in the case of liquid reflective material) a reflective material. The structural material or materials used in a reflector 108 may be substantially neutronically translucent to the extent possible, at least on the side facing the reactor core. For example, a reflector 108 may be formed as a liner or vessel of one or more refractory alloys, nickel alloys, carbides, or graphite compounds. For instance, the material used to form the structural components of a reflector 108 may include, but are not limited to, any one or more components or combinations of one or more molybdenum alloys (e.g., TZM alloy), one or more tungsten alloys, one or more tantalum alloys, one or more niobium alloys, one or more rhenium alloys, one or more nickel alloys, silicon carbide, or graphite compounds, and the like. In an alternative embodiment, a neutron shield (not shown in FIG. 1) may be provided to reduce the neutron exposure of the components outside of the reactor core vessel 104. For example, as discussed in greater detail with reference to FIG. 2, a shield may be provided between the reactor core 104 and the heat exchangers 110, 112 instead of, or in addition to, a reflector 108 between the reactor core 104 and the heat exchangers 110, 112. One effect of the shield is to reduce the neutron flux through the lower tube sheets and, more generally, to reduce the exposure of the components above the shield to neutrons emitted from the core. Depending on the design, a neutron shield may or may not also act as a reflector 108 or a neutron absorber. FIG. 1 generally introduces the main components of the vertically-segmented reactor. Broadly speaking, the balance of this disclosure describes variations and alternative component configurations for the reactor 100. FIG. 2 illustrates a vertical cross-section of an alternative embodiment of a vertically-segmented reactor. In the reactor 200 shown, a cylindrical reactor core 202 is surrounded on the bottom and sides by reflector material 208. The reactor core is separated from the heat exchangers by a shield 203 at the top of the core 202. The shield 203 is provided with flow channels (not shown) through which the fuel salt can travel between the reactor core and the heat exchangers. In an embodiment, the flow channels are angled so that the neutron flux through the flow channels is reduced. For example, in an embodiment, the flow channels are curved, helically shaped or provided with one or more bends. In an alternative embodiment, a flow channel for the fuel salt that also reduces neutron flux through the flow channel may be created by filling the flow channel with a loose, random packing material, such as pellets, tubes, pebbles, saddles, spheres, or rings, of the shield material. In this embodiment, the flow channel essentially acts as a packed bed filter that allows the fuel salt to pass but that intercepts some or all of the neutrons passing through the flow channel. While these configurations increase the resistance to fluid flow, the neutron flux through the channels is reduced by eliminating any straight, unshielded path between the reactor core and the heat exchangers. One possible shield 203 could be a sheet or frame of structural steel upon which a layer of absorber material is connected. Examples of other materials suitable for use in a shield 203 include boron, boron carbide, and some rare earth elements. In the reactor 200 shown, the first and second heat exchanger stages are provided as a single, integrated heat exchanger assembly 210. The integrated heat exchanger assembly 210 includes a plurality of vertically oriented tubes 222A, 222B within a single shell 224 and capped at both ends by a tube sheet 226. In the embodiment shown, fuel salt flows upward through the center tubes 222A of the integrated heat exchanger assembly 210, which perform the function of the first heat exchanger stage. The fuel salt exits the central tubes 222A into a pump plenum 228 located above the integrated heat exchanger assembly 210. Pump impellers 230 located in the plenum 228 circulate the fuel salt into the exterior tubes 222B of the integrated heat exchanger assembly 210 where it flows downward and back into the reactor core. To actively drive the flow through the components of fuel circuit, the reactor 200 may have multiple impellers 230 (as illustrated), a centrally located axial impeller that drives the fuel salt laterally to the periphery of the plenum, or a single, large “waterwheel” impeller 230 that rotates within the plenum about the center axis and drives flow down into the second heat exchanger stage's tubes. Fuel salt flow in the plenum 228 is illustrated by the dashed arrows. Baffles for flow control within the plenum 228 may be provided to assist in routing the coolant. In an alternative embodiment, instead of being an open space, some or all of the plenum 228 may be replaced by pipes, ducts, or channels formed in a solid element that acts as a manifold. For example, FIG. 3 illustrates one possible plenum 228 configuration. In this configuration, the cooled fuel salt is delivered into the reactor core 202 from the exterior tubes 222B at the periphery of the core as shown. Baffles 232 are provided in the core to assist in directing the flow of the fuel salt through the reactor 200. Fuel salt flow in the reactor core is illustrated by the dashed arrows. Baffles 232 may be provided in any form or shape in order to achieve any desired flow profile, assist in mixing the fuel salt, or prevent flow dead spots within the core 202. In the embodiment shown, primary coolant flows into the bottom of the shell 224 of the integrated heat exchanger assembly 210 through a coolant inlet 234 and exits the top of the shell via a coolant outlet 236. However, any number of coolant inlets and outlets 234, 236 at any location around the shell 224 may be used. Baffles for flow control and for separating regions within the shell 224 may be provided to control the exchange of heat between the fuel salt and the coolant. The reactor 200 design is particularly suited for a circular horizontal cross-section. Any other desired horizontal cross-sectional shape may be used, such as ellipsoidal, hexagonal, rectangular, square, octagonal, triangular, etc. Individual non-circular horizontal cross-sections may not be as efficient as circular cross-sections in their production of power relative to the amount of fuel salt required, but may provide other advantages such as when packing multiple, independent reactors 200 together in a single containment vessel. In yet another embodiment (not shown), the integrated heat exchanger assembly 210 may be replaced by heat exchanger tubes that flow through a “pool” of coolant. Primarily this embodiment differs from the integrated heat exchanger assembly 210 embodiment of FIG. 2 in that the outer shell 224 of the heat exchanger defines the bounds of a pool of coolant. Coolant may be injected into the pool and extracted from the pool at any one or more locations in the pool to obtain good heat transfer between the heat exchanger tubes and the coolant. FIGS. 4A-D illustrate examples of some of the many possible horizontal cross-sections A-A of the integrated heat exchanger assembly 210 for the vertically-segmented reactor of FIG. 2. Note that in FIGS. 4A-D, the first heat exchanger stage 210A and the second heat exchanger stage 210B are defined by the flow in the tubes in those portions. Although a dashed line is provided for convenience to illustrate the general horizontal shape, in an embodiment the dashed line may or may not be a physical structure such as a baffle or internal shell. The first heat exchanger stage 210A is that area of the integrated heat exchanger assembly 210 that contains the tubes 222A through which hot fuel salt leaves the reactor core 202. These tubes 222A are illustrated with diagonal cross hatching. The second heat exchanger stage 210B is that portion through which the fuel salt flows through the tubes 222B downward before returning into the reactor core 202. The tubes 222B with downward flow being illustrated with a horizontal striped fill. FIG. 4A illustrates a simple embodiment in which the cross-section of the first heat exchanger stage 210A is circular and the second heat exchanger stage 210B is an annulus centered on the first heat exchanger stage 210A. FIG. 4B illustrates another configuration in which the integrated heat exchanger assembly 210 is annular in shape centered on a central region 250. In this configuration, both the first heat exchanger stage 210A and the second heat exchanger stage 210B are annular in shape centered around the central region 250. In one embodiment, the central region 250 is not part of the heat exchanger 210. In this embodiment the central region 250 may be used to provide access to the reactor core 202 for instruments, control rods, and other ancillary equipment (not shown). In an alternative embodiment, some or the entire central region may be used for flowing primary coolant into or out of the first heat exchanger stage 210A. FIG. 4C is an embodiment in which the integrated heat exchanger assembly 210 is square in cross-section, but the first heat exchanger stage is circular. This illustrates that the first heat exchanger stage 210A and the second heat exchanger stage 210B need not be the same in cross-sectional shape, nor do they need to be centered or nested even though those embodiments are more commonly illustrated herein. FIG. 4D illustrates yet another embodiment in which the integrated heat exchanger assembly 210 is circular in cross-section but separated into two semi-circular halves: one for the first heat exchanger stage 210A and one for the second heat exchanger stage 210B. In this embodiment, the lower tube sheet 226 and/or the shield may act as a manifold directing hot fuel salt into the first heat exchanger stage 210A from the center of the reactor core 202 and returning cooled fuel salt to the reactor core 202 from the downcomer tubes 222B at the periphery of the core 202 as shown in FIG. 2. FIGS. 5A-D illustrate an alternative embodiment of a vertically-segmented reactor 500 utilizing a first heat exchanger stage 510 and eight, independent second heat exchanger stages 512. In the embodiment shown, the vertically-segmented molten fuel reactor 500 includes a reactor core vessel 504 defining a reactor core 502 of a fissionable fuel salt that is liquid at the operating temperature within a containment vessel 518A. The reactor core 502 is capped with a shield 503 and otherwise surrounded by reflector material 508. A lone first heat exchanger stage 510 is provided. In the embodiment shown, the first heat exchanger stage 510 is a single pass, circular, shell-and-tube heat exchanger located above the central region of the reactor core 502. In the embodiment shown, the tubes of the first heat exchanger stage 510 penetrate the shield 503 to allow flow from the core 502. Eight, independent second heat exchanger stages 512 are provided in a ring around the first heat exchanger stage 510. In the embodiment shown, each second heat exchanger stage 512 is a single pass, circular, shell-and-tube heat exchanger located above the central region of the reactor core 502. At the top of the first heat exchanger stage 510 is a manifold 511 that distributes the fuel salt to each of eight, independent second heat exchanger stages 512. Eight U-shaped pipes 548 connect the manifold 511 to a pump plenum 528 above each of the eight second heat exchanger stages 512. An expansion tank 552 is located above the manifold 511 that protects the fuel circuit from over pressure conditions due the expansion of the fuel salt during operation. An impeller 530 is located in each of the pump plenums 528, each impeller 530 is provided with a shaft 531 that extends upwards that is driven by a motor or other equipment (not shown) above the level of the pump plenums 528. Such motors may be within the reactor 500, for example located near the top of the containment vessel cap 518B, or external to the reactor 500 with the shafts penetrating the containment vessel cap 518B. Coolant flow is similar to that described with reference to FIG. 1. Cold coolant enters each of the second heat exchanger stages 512 at a cold coolant inlet 534 via a cold coolant delivery pipe 535 which comes in from the top of the containment vessel 518. The coolant is routed via interior baffles 538 through the second heat exchanger stages 512 to a second heat exchanger stages coolant outlet 536, which delivers the coolant directly to one of the eight first heat exchanger stage's coolant inlets 540. In the embodiment shown, coolant flows through the first heat exchanger stage 510, again routed by baffles to improve heat transfer, and exits the first heat exchanger stage 510 via one of the eight first heat exchanger stage coolant outlets 542, to be removed from the reactor 500 via hot coolant removal pipes 544. Flow of fuel salt through the core 502 is again illustrated by the dashed arrows. Cooled fuel salt is returned at the periphery of the reactor core 502 and directed by a baffle 532 and also by a roughly conically-shaped contour 546 provided in the base of the reactor vessel 504. FIG. 5B is a perspective view of a cross-section of the reactor 500 showing more detail regarding the first and second heat exchanger stages 510 and 512 and their interconnections. In particular, FIG. 5B shows more detail regarding the expansion tank 552 connected to the manifold above the first heat exchanger stage 510. FIG. 5C is a plan view looking down on the reactor 500 with the containment vessel cap 518B removed. FIG. 5D is a horizontal cross-section along section A-A′ of FIG. 5A showing further details of the upper tube sheets 526 of the heat exchangers (the upper tube sheets 526 of the second heat exchanger stages are partially obscured by the blades of the impellers 530) and locations of the impellers 530 and coolant piping 535, 544. FIGS. 6A-D illustrate yet another embodiment of the vertically-segmented reactor. In the reactor 600 shown, four first heat exchanger stages 610 are provided above the central region of the reactor core 602. Each first heat exchanger stage 610 is connected to and forms a coolant circuit with one of the four second heat exchanger stages 612 and is connected to and delivers fuel salt to a different one of the four second heat exchanger stages 612. In other aspects, the reactor 600 is similar to that described with reference to FIG. 5. FIG. 6A is a vertical cross-section of the reactor 600 taken along the section B-B′ identified in FIG. 6B and illustrates the fuel connections between first and second heat exchanger stages 610, 612. The cross-section is cut through one pair of first and second heat exchanger stages 610, 612 that are in the same fuel circuit to illustrate the connections and components in the fuel circuit. With regards to the fuel circuit, each first heat exchanger stage 610 is connected to the inlet of a second heat exchanger stage via a U-shaped pipe 611 containing an impeller 630. Again, flow of fuel salt through the fuel circuit is illustrated with dashed arrows. FIG. 6B is a horizontal cross-section along section A-A′ of FIG. 6A showing how the four primary and four second heat exchanger stages 610, 612 are arranged above the reactor core and illustrate the connections and components in the coolant circuit. FIG. 6B shows details regarding the locations of the coolant inlets and outlets. In particular, in the embodiment shown cold coolant is delivered to a coolant inlet 634 of each second heat exchanger stage 612. The coolant outlet 636 of the second heat exchanger stage is connected to a coolant inlet 640 of its associated first heat exchanger stage 610. In the embodiment shown, coolant is removed from each first heat exchanger stage 610 via a first heat exchanger stage coolant outlet 642. In an alternative embodiment, fuel salt may distributed by a manifold to multiple second heat exchanger stages 612, which allows a vertically-segmented reactor to have a different number of first heat exchanger stages 610 than second heat exchanger stages 612. FIG. 6C is a perspective view of the components located above the shield further illustrating the arrangement of the four sets of first and second heat exchanger stages. In the embodiment shown, each heat exchanger is the same size which improves the ease of repair and servicing. FIG. 6D is a partially transparent plan view of the vertically-segmented reactor that shows the locations of the first heat exchanger stages 610 and second heat exchanger stages 612 relative to the baffle 632 and the side wall of the reactor core 602. FIG. 7 is a simplified flow diagram of the cooling process as performed by the vertically-segmented reactor. The method starts with a setup operation 702. The setup operation 702 may or may not include building an embodiment of a vertically-segmented reactor, as described above, but does include providing at least a reactor core containing molten nuclear fuel. As described above, the neutronics of the reactor core cause a nuclear reaction in the nuclear fuel that causes the nuclear fuel to increase in temperature. By the nature of the reaction, the temperature will not be evenly distributed throughout the core with the nuclear fuel in the center of reactor core becoming hotter than the fuel at the bottom or periphery of the reactor core. The high temperature nuclear fuel is then displaced from the reactor core by delivering lower temperature nuclear fuel to the reactor core in a displacement operation 704. In an embodiment, the temperature difference between the high temperature nuclear fuel and the lower temperature nuclear fuel is from 100 to 1000° C., depending on the fuel salt. The greater the temperature difference, the better from a heat transfer perspective. However, certain fuel salts may be very corrosive or otherwise require very expensive equipment to handle at high or very high temperatures. Additionally, the lower temperature is limited by the melting point of the type of nuclear fuel chosen. For example, in one embodiment, the nuclear fuel is 71 mol % UCl4-17 mol % UCl3-12 mol % NaCl and the temperature difference is from 200 to 400° C., with a temperature difference of 250-350° C. being desired for a particular reactor configuration. In an embodiment, the displacement operation 704 is an ongoing operation that maintains a continuous flow of nuclear fuel around the fuel circuit of a vertically-segmented reactor described above. However, during initial reactor start up the displacement operation 704 may include initiating the circulation. In an embodiment in which the fuel circuit is first filled with nuclear fuel, the circulation may be self-initiating by the creation of a natural circulation cell as a result of the temperature in the reactor core increasing relative to the nuclear fuel in the rest of the fuel circuit, upon removal of control rods, for example. As mentioned above, higher temperature molten nuclear fuel is less dense than low temperature nuclear fuel. This density difference creates a buoyancy force that naturally drives the higher temperature nuclear fuel upward at the center of the reactor core and into the first heat exchanger stage, thus initiating natural circulation. In an alternative embodiment, the displacement may be actively initiated through the use of one or more impellers as provided in some of the reactor embodiments prior to removal of the control rods. In this embodiment, upon establishment of circulation in the fuel circuit and criticality in the reactor core, the impellers may be disengaged, stopped, or allowed to freewheel in favor of allowing natural circulation within the fuel circuit. The method 700 further includes a cooling operation 706 in which coolant is routed through the coolant circuit of the heat exchangers to remove heat from the displaced nuclear fuel. The term ‘routed’ is used as the flow of coolant may be either actively maintained via pumping or passively maintained via natural circulation. In an embodiment, while in steady state operation the flow of coolant through the coolant circuit may also be driven primarily or completely by natural circulation due to the heating of the coolant as it passes through the coolant circuit. While FIG. 7 illustrates the cooling method 700 during steady state operation, depending on the embodiment, the cooling operation 706 may be initiated before, concurrently or after initiation of the displacement operation 704. In an embodiment, the temperature of the coolant delivered to the heat exchangers is at or below that of the low temperature nuclear fuel. In an embodiment, coolant is routed in a coolant circuit first through the second heat exchanger stage(s) and then through the first heat exchanger stage(s) as described above. Alternatively, each heat exchanger may be a separate independent coolant circuit. Regardless of the coolant circuit configuration, as part of the cooling operation 706 the temperature of the coolant delivered to the coolant circuit may be actively controlled to achieve a target operational parameter. For example, the temperature of the coolant delivered to the reactor could be controlled to maintain a target steady state reactor core temperature, a target heat removal rate, a target temperature for the low temperature nuclear fuel, and/or based on any other operational parameter of the reactor. The method 700 may further include neutronically shielding the first and second heat exchangers from neutrons generated in the reactor core. As described above this may be achieved passively by providing a neutron shield between the reactor core and the heat exchangers. FIG. 8 illustrates an embodiment of a reflector configuration that may be used in a vertically-segmented reactor similar to that of FIG. 1. In the embodiment shown, individual reflectors in the form of structural tubes 808 filled with reflector material, which may be solid or liquid at operating temperatures, are provided. The reflector tubes 808 may be located within the lower vessel portion 818A of the containment vessel but outside of the reactor core vessel 804 (as illustrated) or inside the reactor core vessel (e.g., in contact with the nuclear fuel) or both. For example, in the embodiment illustrated the reflector tubes 808 are provided in two concentric rings around the sides of the reactor core vessel 804. A shield 803 is provided to protect the heat exchangers 810, 812, pump 814 and other components above the core 802 within the containment vessel 818 from neutron damage. An additional reflector 890, which may or may not be in the form of a tube, may also be provided below the reactor core 802 as shown. The number and arrangement of the reflector tubes 808 are selected to provide the desired amount of reflection back into the core of neutrons that would otherwise be lost to the core 802. In an embodiment, the tubes 808 may be collected and formed into a tube bundle. In an alternative embodiment, each tube may be independent and unconnected to the other tubes in the reactor, allowing individual tubes to be replaced easily. One or more of the tubes 808 may be movable in order to dynamically alter the neutron flux in the core 802. In the embodiment shown, the reflector tubes are provided with a connecting rod 809 that allows the tubes to be raised and lowered from above. The connecting rods 809 also allow the reflecting tubes 808 to be easily inspected and replaced, if necessary, by lifting them out of the top portion 818B of the containment vessel 818. The capability to remove a reflector tube from the reactor further allows flexibility in operation as a removed reflector tube may be replaced with a control rod or an instrument for obtaining information from the reactor core 802. In an alternative embodiment, one or more of the tubes in the reflector tube bundle may be dedicated to use as an instrument-containing tube or control rod. The reflector tubes 808 in this configuration may be held within a structural framework (not shown) to maintain proper alignment or may simply be hanging from the connecting rods. A structural framework could be a block of solid material provided with passages for reflector tubes or could be an open, lattice structure. FIG. 9 illustrates an embodiment of yet another reactor that uses reflector tubes within the reactor core vessel. This configuration has the benefit of reducing neutron damage to the reactor core vessel 904 by reflecting at least some neutrons before they reach the vessel 904. Otherwise, this embodiment is similar to that of FIG. 8 in that the reflector tubes 908 are provided with a connecting rod 909 that allows the tubes to be raised and lowered from above. In the embodiment shown, space is provided around the reflector tubes to allow fuel salt to circulate between the tubes 908. It may be desirable to maintain the reflector tubes 908 at a lower temperature and flow of the low temperature fuel salt entering the reactor core from the second heat exchanger stage may be partially or completely directed through or around the reflector tubes 908 to maintain them at a lower temperature than that of the central core during operation. Again, a framework (not shown) may be provided to hold the tubes 908 in position and may also have flow channels for directing the low temperature fuel salt flow around the tubes. FIG. 10 illustrates yet another embodiment of a framework and configuration of reflector tubes in a vertically-segmented reactor that prevents stagnation of the fuel salt within the reflector tubes and provides more active cooling of the reflector tubes by directing the flow of the cold fuel salt exiting the second heat exchanger stages through the reflector tube bundle before it enters the center of the reactor core. In the reactor 1000, fuel salt from the reactor core 1002 flows upward through a first heat exchanger stage 1010 and then is routed to peripheral second heat exchanger stages 1012 where the fuel flows downward back into the reactor vessel 1004. As with FIG. 9, the reactor 1000 includes reflector tubes 1008 at the periphery of the reactor core 1002 within the reactor vessel 1004. However, the reactor 1000 is provided with one or more baffles 1032 that direct the flow of the cooled fuel salt past the reflector tubes before the fuel salt flows into the central region of the reactor core 1002. In the embodiment shown, the baffles 1032 form a solid wall between the reflector tubes 1008 and the central region of the reactor core 1002, with the exception of a perforated zone 1033 near the bottom of the vessel 1004. This directs the flow of the cold fuel salt as illustrated by the dashed flow arrows and ensures that the reflector tubes 1008 remain at a temperature close to that of the cold fuel salt exiting the second heat exchanger stages 1012. While other configurations are also possible including FIG. 16, FIG. 10 illustrates that the cooled fuel salt can be used to actively cool the reflector and, indeed, any components located in the reactor core 1002 or near a surface of the reactor vessel. For example, in yet another embodiment, the baffle 1032 could be in a U shape that further forces the cooled fuel salt to flow along the bottom of the reactor vessel 1004 before entering the central region of the reactor core 1002 through a perforated zone at the center of the U. This configuration would keep the lateral reflector tubes as well as the bottom reflector 1090 cool during operation. FIG. 16 illustrates an alternative embodiment of a vertically-segmented reactor that provides for a more active cooling of the reflector tubes, in this case by diverting a portion of the flow of the cold fuel salt exiting the second heat exchanger stages through the reflector tube bundle before it enters the center of the reactor core 1602 while the balance of the flow is directed into the reactor core. In the reactor 1600, fuel salt from the reactor core 1602 flows upward through a first heat exchanger stage 1610 and then is routed to peripheral second heat exchanger stages 1612 where the fuel flows downward back into the reactor vessel 1604. As with FIGS. 9 and 10, the reactor 1600 includes reflector tubes 1608 at the periphery of the reactor core 1602 within the reactor vessel 1604. Flow from the second heat exchanger stages 1612 is split, for example as it passes through the shield 1603 as shown, so that a portion of the flow is directed to the reflector tubes 1608 and the balance flows into the center of the reactor core 1602. As in FIG. 10, the reactor 1600 is provided with one or more baffles 1632 that directs the flow of cooled fuel salt past the reflector tubes before that portion of the fuel salt flows into the central region of the reactor core 1602. In the embodiment shown, the baffles 1632 form a solid wall between the reflector tubes 1608 and the central region of the reactor core 1602, with the exception of a perforated zone 1633 near the bottom of the vessel 1604. This configuration directs the flow of the cold fuel salt as illustrated by the dashed flow arrows. In an embodiment, the amount of flow split between the reflector tubes 1608 and that directly entering the central region of the reactor core 1602 is controlled by the design of the shield. In an embodiment, 10-30%, e.g., 20%, of the flow may be diverted past the reflector tubes with the balance delivered to the central region of the reactor core 1602. In an alternative embodiment, the flow diversion may be adjustable and actively controlled by moveable valves, baffles, or other flow diversion equipment provided in the fuel circuit, such as in the shield 1603 or at the bottom of the second heat exchanger stage 1612. FIG. 16 also illustrates the use of a downcomer 1688, or dip tube, to deliver the cooled fuel salt into the bottom of the reactor core 1602 as shown. This may assist with creating a strong natural circulation cell in the fuel circuit and may also prevent hot spots or areas of low circulation from being created in the reactor core 1602. FIG. 11 illustrates an embodiment of a framework and configuration of reflector tubes in a vertically-segmented reactor. FIG. 11 illustrates a vertical cross-section of a portion of the reactor core vessel sidewall 1104, the reactor core vessel floor 1105, and a shield 1103. Section lines A-A′, B-B′, C-C′ and D-D′ are indicated on the cross-section and, right of the section lines, a horizontal cross-section is shown associated with each section line to indicate the relative arrangement of the reflector tubes at that point in the reactor. In the embodiment shown, some reflector tubes 1108A are contained completely within the reactor core vessel sidewall 1104. Other reflector tubes 1108B are exposed to the fuel salt in the reactor core 1102. In the embodiment shown, the reflector tubes 1108A, 1108B also penetrate at least some distance into the reactor core vessel floor 1105. A ramp for directing fuel salt flow is provided as shown at section C-C′. Reflector tubes of FIGS. 8-11 may be of any cross-sectional shape, such as for example circular (i.e., cylindrical tubes), hexagonal, octagonal, ellipsoidal, etc. In addition reflector tubes may have different external and internal profiles, such as for example a hexagonal exterior cross-section with a circular interior space for the reflector material. Tubes may be contoured for strength, for directing flow of fuel salt around the exterior or for positive engagement with a framework. A Direct Reactor Auxiliary Cooling System (DRACS) independent of the power generating heat exchanger circuit is often used to enhance the safety of the reactor. The embodiments of the vertically-segmented reactor described above could be easily adapted to include additional DRACS heat exchangers in the fuel circuit, such as above the first and second heat exchanger stages. This addition would not require increasing the size of the reactor vessel or reactor core and, thus, the vertically-segmented reactor is well-suited for use with DRACS. FIGS. 12-14 illustrate how several of the reactor embodiments described above could be adapted to use a DRACS. As an independent heat exchanger circuit, one or more DRACS heat exchangers 1202 could be located anywhere in the fuel circuit of a vertically-segmented reactor. FIG. 12 illustrates a reactor 1200 similar to that illustrated in FIG. 1, in which the DRACS heat exchanger is located at the top of the second heat exchanger stage 112. A separate DRACS cooler 1204 and coolant circuit is provided that brings coolant into the DRACS heat exchanger 1102 and returns it to the cooler 1204. Alternative embodiments include locating the DRACS heat exchanger 1202 at the bottom of the second heat exchanger stage 112, above the first heat exchanger stage 110, between the first and second heat exchanger stages, or below the first heat exchanger stage 110. However, it may be beneficial in creating strong natural circulation to locate the DRACS heat exchanger as far above the center of the reactor core 102 as possible. FIG. 13 illustrates how a DRACS could be integrated into a vertically-segmented reactor as shown in FIG. 2. In this embodiment, a DRACS heat exchanger 1302 is placed between the integrated heat exchanger assembly 210 and the pump plenum 228 of the reactor 1300. In an alternative embodiment, the DRACS heat exchanger could be between the bottom of the integrated heat exchanger assembly 210 and the plenum 228. In yet another embodiment the DRACS heat exchanger 1302 could be integrated into the integrated heat exchanger assembly 210 by using the same heat exchanger tubes 222A, 222B of the integrated heat exchanger, but isolating that portion in the shell so that an independent DRACS coolant in used in that portion. FIG. 14 illustrates how a DRACS could be integrated into a vertically-segmented reactor as shown in FIGS. 5A-D. In this embodiment, a DRACS heat exchanger 1402 is placed in each of four of the U-shaped pipes 548 in the reactor 1400. In an alternative embodiment, fewer or more of the U-shaped connecting pipes may be provided with a DRACS heat exchanger 1402 to achieve the desired amount of cooling. FIGS. 12-14 illustrate the flexibility of the vertically-segmented reactor for locating DRACS heat exchangers in the fuel circuit. FIG. 15 illustrates how a vertically-segmented reactor may be adapted for use with a coolant pool. In the reactor 1500 shown, the containment vessel 1518 is filled with coolant to form a pool within which the reflector 1508, reactor vessel 1504, first heat exchanger stage 1510 and second heat exchanger stage 1512 are submerged. In the pool embodiment, the first heat exchanger stage 1510 and second heat exchanger stage 1512, instead of being shell-and-tube heat exchangers, take the form of tube bundles, essentially a shell-and-tube exchanger with the shell removed. Thus, the coolant is free to flow between the tubes of the first heat exchanger stage 1510 and second heat exchanger stage 1512 to provide cooling to the fuel salt within the tubes. Fuel salt may be actively flowed using a pump 1514 that may or may not be below the level of the coolant in the pool, or the flow may be driven by natural circulation. Likewise, the coolant may be actively circulated within the pool, such as by driving the coolant flow from the second heat exchanger stage 1512 toward the tubes of the first heat exchanger stage 1510 before coolant is removed. Coolant flow may be directed by the injection point of the coolant, baffles, and/or impellers located at points within the pool to obtain the desired flow profile around the submerged components. This configuration allows the coolant to further provide cooling to the reflector 1508 at the sides and the bottom of the reactor vessel 1504. Notwithstanding the appended claims, the disclosure is also defined by the following clauses: 1. A molten fuel nuclear reactor comprising: a reactor vessel defining a reactor core containing nuclear fuel; a first heat exchanger above the reactor core that receives high temperature nuclear fuel from the reactor core; a second heat exchanger above the reactor core that receives nuclear fuel from the first heat exchanger and delivers lower temperature nuclear fuel to the reactor core; and a containment vessel surrounding the reactor vessel, the first heat exchanger, and the second heat exchanger. 2. The molten fuel nuclear reactor of clause 1 further comprising: one or more impellers within the containment vessel that drive the flow of fuel through the reactor vessel, the first heat exchanger, and the second heat exchanger. 3. The molten fuel nuclear reactor of clause 1 or 2 further comprising: a neutron shield separating the reactor core from the first and second heat exchangers. 4. The molten fuel nuclear reactor of any of the above clauses further comprising: a reflector assembly surrounding at least a portion of the reactor vessel. 5. The molten fuel nuclear reactor of any of the above clauses further comprising: a reflector assembly within the reactor vessel located within the nuclear fuel at a periphery of the reactor core. 6. The molten fuel nuclear reactor of any of the above clauses further comprising: one or more baffles affecting nuclear fuel flow in at least one of the reactor core, the first heat exchanger, and the second heat exchanger. 7. The molten fuel nuclear reactor of any of the above clauses further comprising: a plenum between a nuclear fuel outlet of the first heat exchanger and a nuclear fuel inlet of the second heat exchanger. 8. The molten fuel nuclear reactor of any of the above clauses, wherein during operation natural circulation drives the flow of nuclear fuel through the reactor vessel, the first heat exchanger, and the second heat exchanger, the natural circulation created by a temperature difference between high temperature fuel in the reactor core and the lower temperature fuel exiting the second heat exchanger. 9. The molten fuel nuclear reactor of any of the above clauses, wherein the nuclear fuel in the reactor core is a salt of chloride, bromide, and/or fluoride. 10. The molten fuel nuclear reactor of any of the above clauses, wherein the nuclear fuel contains one or more of uranium, plutonium, or thorium. 11. The molten fuel nuclear reactor of any of the above clauses, wherein the first heat exchanger and the second heat exchanger are contained within a single shell. 12. The molten fuel nuclear reactor of any of the above clauses wherein the first heat exchanger is a single, shell-and-tube heat exchanger. 13. The molten fuel nuclear reactor of any of the above clauses, wherein one or both of the first heat exchanger and the second heat exchanger includes one or more individual, shell-and-tube heat exchangers. 14. A method for removing heat from a molten fuel nuclear reactor having a reactor core containing high temperature liquid nuclear fuel, the method comprising: delivering low temperature nuclear fuel into the reactor core, thereby displacing some high temperature nuclear fuel from the reactor core upward through a first heat exchanger and downward through a second heat exchanger; and routing coolant through the first and second heat exchangers, thereby transferring heat from the high temperature nuclear fuel to the coolant and converting the displaced high temperature nuclear fuel into the low temperature nuclear fuel. 15. The method of clause 14, wherein delivering the low temperature nuclear fuel into the reactor core includes passing the low temperature nuclear fuel from the second heat exchanger into the reactor core. 16. The method of clause 14 or 15, wherein delivering the low temperature nuclear fuel includes operating at least one impeller to drive flow of the nuclear fuel through the first and second heat exchangers. 17. The method of any of clauses 14-16 further comprising: neutronically shielding the first and second heat exchangers from neutrons generated in the reactor core. 18. The method of any of clauses 14-17, wherein routing the coolant includes delivering coolant at a temperature less than that of the high temperature nuclear fuel to the second heat exchanger. 19. The method of any of clauses 14-18, wherein routing coolant includes pumping coolant first through the second heat exchanger and then through the first heat exchanger. 20. The method of any of clauses 14-19, wherein the first and second heat exchangers are vertically-oriented shell-and-tube heat exchangers located above the reactor core. Notwithstanding that the numerical ranges and parameters setting forth the broad scope of the technology are approximations, the numerical values set forth in the specific examples are reported as precisely as possible. Any numerical values, however, inherently contain certain errors necessarily resulting from the standard deviation found in their respective testing measurements. It will be clear that the systems and methods described herein are well adapted to attain the ends and advantages mentioned as well as those inherent therein. Those skilled in the art will recognize that the methods and systems within this specification may be implemented in many manners and as such are not to be limited by the foregoing exemplified embodiments and examples. In this regard, any number of the features of the different embodiments described herein may be combined into one single embodiment and alternate embodiments having fewer than or more than all of the features herein described are possible. While various embodiments have been described for purposes of this disclosure, various changes and modifications may be made which are well within the scope contemplated by the present disclosure. For example, in any of the embodiments shown, the positions of the first heat exchanger stages and second heat exchanger stages may be reversed so that cooled fuel salt enters the reactor directly above the center of reactor core and hot fuel salt is removed at the periphery of the reactor core. Such a modification only requires reversing the direction of flow in the fuel circuit. As another example, when adapted for use in thermal reactors the embodiments of the vertically-segmented reactor may include one or more moderators to thermalize the neutrons in the reactor core. Such moderators may be located in the reactor core and may be components made with or including graphite, water, beryllium, or beryllium oxide. Numerous other changes may be made which will readily suggest themselves to those skilled in the art and which are encompassed in the spirit of the disclosure.
	description	1. Technical Field The present disclosure relates to a protective clothing, and more particularly, to a radiation resistant clothing for resisting electromagnetic radiation (EMR). 2. Description of Related Art Protective clothing, such as radiation resistant clothing, is necessary for pregnant women, children, hospital patients, and some workers exposed to hazardous EMR. Radiation resistant clothing usually has metal fibers for reflecting EMR away from a body of a person. However, some radiation can still reach the body via sleeves and neckline, etc., and when the radiation reaches into a space between the body and the clothing, it will be reflected to the body by the metal fibers of the clothing, such that the body may still absorb significant radiation. What is needed is to provide a radiation resistant clothing that can overcome the above-described limitations. Reference will be made to the drawings to describe certain exemplary embodiments of the present disclosure. Referring to FIG. 1 and FIG. 2, an article of radiation resistant clothing 10 is shown. The radiation resistant clothing 10 includes an outer cloth layer 111, a first radiation resistant layer 113 for reflecting radiation, a second radiation resistant layer 115 for absorbing radiation, and an inner cloth layer 117. The first radiation resistant layer 113 and the second radiation resistant layer 115 are sandwiched between the outer cloth layer 111 and the inner cloth layer 117. The second radiation resistant layer 115 is positioned on the inside of the first radiation resistant layer 113 and next to the inner cloth layer 117. The first radiation resistant layer 113 includes EMR shielding material for reflecting EMR. In detail, the EMR shielding material can reflect most EMR. In one embodiment, the first radiation resistant layer 113 includes mixed material made of the radiation shielding material and common clothing fiber material. The EMR shielding material may include material selected from the group consisting of metal fiber material and nanometer metal fiber material. The metal fiber material may be stainless steel fiber material or silver fiber material, and the nanometer metal fiber material may be nanometer silver fiber material. The second radiation resistant layer 115 includes EMR absorbing material for absorbing EMR as opposed to reflecting EMR. In detail, the EMR absorbing material converts most EMR into heat or other energy. In one embodiment, the second radiation resistant layer 115 includes mixed material made of the radiation absorbing material and the common clothing fiber material. The EMR absorbing material can be silicon carbide fiber material (such as nanometer silicon carbide fiber material) or multi-ion fabric material (such as multi-ion acrylic fiber material), and the common clothing fiber material can include material selected from the group consisting of bamboo rayon fiber material, bamboo carbon fiber material, cotton fiber material, polyester fiber material, and polyamide fiber material. Both of the outer cloth layer 111 and the inner cloth layer 117 are made of the common clothing fiber material. The common clothing fiber material includes material selected from the group consisting of bamboo rayon fiber material, bamboo carbon fiber material, cotton fiber material, polyester fiber material, and polyamide fiber material. The radiation resistant clothing 10 includes the a first radiation resistant layer 113 and the second radiation resistant layer 115 positioned on the inside of the first radiation resistant layer 113, thus even EMR which penetrates under the clothing through crevices and open gaps, such as via the ends of sleeves and the neckline, is absorbed by the second radiation resistant layer 115 to avoid the greatest possible protection for the body. Accordingly, the protection effect of the radiation resistant clothing 10 is greatly improved. Referring to FIG. 3, a side view of material of the radiation resistant clothing 20 of a second embodiment of the present disclosure is shown. The radiation resistant clothing 20 differs from the radiation resistant clothing 10 only in that the inner cloth layer 117 is omitted, but an outer layer 211, a first radiation resistant layer 213 and a second radiation resistant layer 215 are respectively the same as the outer layer 111, the first radiation resistant layer 113 and the second radiation resistant layer 115. The second radiation resistant layer 215 includes mixed materials made of radiation absorbing material and common clothing fiber material and serves as an inner layer of the radiation resistant clothing 20. Referring to FIG. 4, a side view of material of the radiation resistant clothing of a third embodiment of the present disclosure is shown. The radiation resistant clothing 30 differs from the radiation resistant clothing 20 only in that the outer layer 111 and the inner cloth layer 117 are omitted, but a first radiation resistant layer 313 and a second radiation resistant layer 315 are respectively the same as the first radiation resistant layer 213 and the second radiation resistant layer 215. The first radiation resistant layer 313 includes mixed materials made of radiation resistant material and common clothing fiber material and serves as an outer layer of the radiation resistant clothing. It is to be further understood that even though numerous characteristics and advantages of preferred and exemplary embodiments have been set out in the foregoing description, together with details of the structures and functions of the embodiments, the disclosure is illustrative only; and changes may be made in detail, especially in the matters of shape, size and arrangement of parts within the principles of the present disclosure to the full extent indicated by the broad general meaning of the terms in which the appended claims are expressed.
050420593	claims	1. An optical element which is adapted for use in a radiation optical system having a radiation source and a means for receiving radiation generated from the radiation source and passing through the optical element, said optical element comprising a graphite film obtained by pyrolyzing a film of a polymer selected from the group consisting of polyphenylene oxadiazole, polybenzothiazole, polybenzobisthiazole, polybenzooxazole, polybenzobisoxazole, polypyromellitimide, polyphenylene-isophthalamide, polyphenylene-benzoimidazole, polyphenylene-benzobisimidazole, polythiazole and poly-p-phenylene-vinylene at a first temperature of not higher than 2800.degree. C. and at a first pressure of not higher than 20 kg/cm.sup.2, and subsequently at a second temperature higher than 2800.degree. C. and at a second pressure higher than 20 kg/cm.sup.2. 2. An optical element according to claim 1, wherein said polymer is polyphenylene oxadiazole. 3. An optical element according to claim 1, wherein said polymer is polypyromellitimide. 4. An optical element according to claim 1, wherein said graphite film is formed on to a flat substrate. 5. An optical element according to claim 1, wherein said graphite film is formed on a cylindrically curved substrate on the inner surface thereof. 6. An optical element according to claim 1, wherein a plurality of said polymer films are stacked to provide a block of graphite films. 7. An optical element according to claim 1, wherein a plurality of graphite films which have been obtained by pyrolyzing a plurality of polymer films at a temperature of not lower than 2800.degree. C. under a pressure of not lower than 4 kg/cm.sup.2, are combined to provide a graphite block. 8. An optical element which is adapted for use in a radiation optical system having a radiation source and a means for receiving a radiation generated from the radiation source and passing through an optical element, said optical element comprising a graphite film which is obtained by pyrolyzing at a temperature of not lower than 2800.degree. C., in vacuum or in an inert gas, a film of a polymer selected from the group consisting of polyphenylene oxadiazole, polybenzothiazole, polybenzobisthiazole, polybenzooxazole, polybenzobisoxazole, polypyromellitimide, polyphenylene isophthalamide, polyphenylene benzoimidazole, polyphenylene benzobisimidazole, polythiazole and poly-p-phenylene-vinylene and which has a metal halide intercalated therein. 9. An optical element according to claim 8, wherein said metal halide is intercalated in a first-stage condition where said metal halide is present in individual space between any adjacent lattice layers of said graphite film. 10. An optical element according to claim 8, wherein said metal halide is intercalated in a second-stage condition where said metal halide is present in every third space of lattice layers of said graphite film. 11. An optical element according to claim 8, wherein said metal halide is at least one member selected from the group consisting of BCl.sub.3, MgCl.sub.2, AlCl.sub.3, ScCl.sub.3, TiCl.sub.4, CrCl.sub.3, MnCl.sub.2, FeCl.sub.3, CoCl.sub.2, NiCl.sub.2, CuCl.sub.2, ZnCl.sub.2, GaCl.sub.3, YCl.sub.3, NbCl.sub.5, MoCl.sub.5, RhCl.sub.3, PdCl.sub.2, CdCl.sub.2, RuCl.sub.3, ZrCl.sub.4, InCl.sub.3, HfCl.sub.4, TaCl.sub.5, WCl.sub.6, ReCl.sub.4, OsCl.sub.4, PtCl.sub.4, AuCl.sub.3, HgCl.sub.2, TlCl.sub.3, BiCl.sub.4, ICl, IBr, FeCl.sub.2, BF.sub.3, AlBr.sub.3, SiF.sub.4, TiF.sub.4, FeBr.sub.3, CuBr.sub.2, PF.sub.6, GaBr.sub.3, NbF.sub.5, MoF.sub.6, CdBr.sub.2, TaF.sub.6, WF.sub.6, OsF.sub.3, AuBr.sub.3 and TlBr.sub.3. 12. An optical element according to claim 11 wherein said metal halide is CuCl.sub.2. 13. An optical element according to claim 11, wherein said metal halide is NiCl.sub.2. 14. An optical element according to claim 8, wherein a plurality of the intercalated graphite films are bonded together to form a sheet or block. 15. An optical element according to claim 8, wherein the intercalated graphite film is formed on a flat substrate. 16. An optical element according to claim 8, wherein intercalated graphite film is formed on a curved substrate at the inner side thereof. 17. An optical element which is adapted for use in a radiation optical system having a radiation source and a means for receiving a radiation generated from the radiation source and passing through an optical element, said optical element comprising at least one graphite film and at least one graphite film intercalated with a metal halide, both films being bonded together by pressing. 18. An optical element according to claim 17, wherein a plurality of the graphite films and a plurality of the intercalated graphite films which are alternately superposed and bonded together by pressing to form a sheet so that the graphite films are placed as an outermost layer on opposite sides. 19. An optical element according to claim 17, wherein a plurality of intercalated graphite films are sandwiched between two graphite films and bonded together by pressing to form a sheet. 20. An optical element according to claim 17, wherein said at least one graphite film is obtained by pyrolyzing at a temperature of not lower than 2800.degree. C. at a normal pressure in vacuum or in an inert gas a film of a polymer selected from the group consisting of polyphenylene oxadiazole, polybenzothiazole, polybenzobisthiazole, polybenzooxazole, polybenzobisoxazole, polypyromellitimide, polyphenylene-isophthalamide, polyphenylene-benzoimidazole, polyphenylene-benzobisimidazole, polythiazole and poly-p-phenylene-vinylene. 21. An optical element according to claim 17, wherein said at least one graphite film is obtained by pyrolyzing at a temperature of not lower than 2800.degree. C. at a pressure of not lower than 4 kg/cm.sup.2 in vacuum or in an inert gas a film of a polymer selected from the group consisting of polyphenylene oxadiazole, polyimide, polybenzothiazole, polybenzobisthiazole, polybenzooxazole, polybenzobisoxazole, polypyromellitimide, polyphenylene-isophthalamide, polyphenylene-benzoimidazole, polyphenylene-benzobisimidazole, polythiazole and poly-p-phenylene-vinylene. 22. An optical element according to claim 17, wherein said at least one graphite film intercalated with the metal halide is a product which is obtained by pyrolyzing at a temperature of not lower than 2800.degree. C. in vacuum or in an inert gas a film of a polymer selected from the group consisting of polyphenylene oxadiazole, polybenzothiazole, polybenzobisthiazole, polybenzooxazole, polybenzobisoxazole, polypyromellitimide, polyphenylene-isophthalamide, polyphenylene-benzoimidazole, polyphenylene-benzobisimidazole, polythiazole and poly-p-phenylene-vinylene and intercalating the resultant graphite film with a metal halide. 23. An optical element according to claim 22, wherein said metal halide is at least one member selected from the group consisting of BCl.sub.3, MgCl.sub.2, AlCl.sub.3, ScCl.sub.3, TiCl.sub.4, CrCl.sub.3, MnCl.sub.2, FeCl.sub.3, CoCl.sub.2, NiCl.sub.2, CuCl.sub.2, ZnCl.sub.2, GaCl.sub.3, YCl.sub.3, NbCl.sub.5, MoCl.sub.5, RhCl.sub.3, PdCl.sub.2, CdCl.sub.2, InCl.sub.3, HfCl.sub.4, TaCl.sub.5, WCl.sub.6, ReCl.sub.4, OsCl.sub.4, PtCl.sub.4, AuCl.sub.3, HgCl.sub.2, TlCl.sub.3 and BiCl.sub.4. 24. An optical element according to claim 17, wherein the bonded films are formed on a flat substrate. 25. An optical element according to claim 17, wherein the bonded films are formed on a cylindrically curved substrate.
043404430	summary	The present invention relates to the determination of the gold content of auriferous materials, and in particular to the determination of the gold content of auriferous rock samples. Gold may occur at depth in thin bands of mineralisation which when mined together are accompanied by substantial quantities of barren rock. In order to prevent the expensive and time consuming treatment of all mined material, it is necessary that some pre-selection process be applied to the mined material. A number of methods of selecting rocks for processing have been proposed but to date no entirely satisfactory method of selection has been found. Some methods have failed because they are secondary methods and the correlation between the secondary property measured and the gold content is either variable or inaccurate; other have not been able to cope with the throughput of samples necessary in a production environment. The present invention provides a method of determining the gold content of mined rock which is both capable of coping with the required throughput of rock samples, and utilises a property of the gold itself to determine its concentration in the rock samples. According to the present invention there is provided a method for determining the gold content of an auriferous material, comprising the operations of irradiating a body of the material with neutrons and determining the intensity of .gamma.-rays having an energy of 279 keV arising from the reaction .sup.197 Au (nn') .sup.197.sbsp.m Au.fwdarw.279 keV. If the method is being used for the determination of the gold content of auriferous rock, then it is necessary to use a neutron source which does not produce neutrons which have an energy above the neutron reaction thresholds of elements such as Al, Si, Ca, Fe and O which are likely to be present in high concentrations. For example, suitable neutron sources are tube sources which utilise the deuteron-deuteron or deuteron-beryllium reaction to produce neutrons.
	summary
047056622	abstract	Fast neutron nuclear reactor of the type comprising a primary circuit integrated into a liquid metal-filled vessel and containing the reactor core, as well as means for circulating the said liquid metal and means for transferring the heat carried by the liquid metal to the water circulating in a water/steam circuit, wherein the heat transfer means comprise at least one steam generator located in the reactor vessel and having at least one group of inner tubes in which circulates the water of the water/steam circuit, at least one group of outer tubes immersed in the primary liquid metal, the outer tube surrounding each of the inner tubes in order to define therewith an annular space under a neutral gas pressure, connected to secondary circuit with a low thermal power having means for circulating this pressurized neutral gas, such as helium in said annular space and ensure the heat exchange.

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