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060312416	abstract	Capillary discharge extreme ultraviolet lamp sources for EUV microlithography and other applications. The invention covers operating conditions for a pulsed capillary discharge lamp for EUVL and other applications such as resist exposure tools, microscopy, interferometry, metrology, biology and pathology. Techniques and processes are described to mitigate against capillary bore erosion, pressure pulse generation, and debris formation in capillary discharge-powered lamps operating in the EUV. Additional materials are described for constructing capillary discharge devices fore EUVL and related applications. Further, lamp designs and configurations are described for lamps using gasses and metal vapors as the radiating species.
041586050	description	DESCRIPTION OF THE PREFERRED EMBODIMENT Referring now to the drawings, there is shown in FIG. 1 a typical nuclear reactor vessel 10 including the vessel head 12. The vessel 10 encloses a reactor core 14 which includes a plurality of elongated fuel assemblies 16 oriented adjacent one another. The assemblies 16 are supported by a lower core plate 18 which is perforated to allow passage of coolant and which in turn is supported by a core barrel 20. The core barrel 20 is supported from a ledge 22 of the reactor vessel 10, and is restrained in lateral movement by a radial support system 24 affixed to the vessel 10. The main flow of reactor coolant fluid typically enters the vessel 10 through one or more inlet nozzles 26, passes downward about the outer periphery of the core barrel 20 and about the affixed neutron shields 28, is turned one hundred and eighty degrees in a lower plenum 30, passes upward through the lower core plate 18 and core 14, and exits through outlet nozzles 32. It is of prime importance that the flow of coolant is carefully controlled into and about the fuel assemblies 16 of the core 14. Baffling of coolant flow about the core 14 has typically been performed by a baffle plates 34 and formers 36 assembly, through which a small bypass flow of reactor coolant is also passed. This bypass flow must be minimized since it decreases the thermal efficiency of the reactor, but must be large enough to adequately cool the surrounding components. In order to also minimize bypass flow between the outermost fuel assemblies 16, which are typically operating at a lower power density than more central assemblies, and the baffle plates 34, the baffle plates 34 are oriented close to the assemblies 16. Because the baffle plates 34, shown best in FIGS. 2 and 3, are typically significantly thinner than the core barrel 20, the differential thermal expansion between them must be accommodated in the means fastening the formers 36 to the barrel 20 and to the baffle plates 34, typically bolts. The differential thermal expansion is compounded not only by the fact that the baffles 34 are closer to the core 14 and the hotter coolant fluid than the core barrel 20, but also because the heat generation in these components changes throughout the reactor operating cycle. Excessive fastener loads are alleviated by this invention, the pasic principle of which is to split the baffle plates 34 transversely at one or more elevations and allow expansion without significant interference at the junction region 38, thereby reducing loadings on the fasteners. To minimally disturb the coolant flow on either side of the baffle plates 34, the upper and lower plates 34 at each junction region 38 should be aligned longitudinally, to present a substantially continuous inner surface 40 and outer surface 42, as shown in FIG. 2a. Further, to minimize any cross flow leakage across the baffle plates 34 at the junction region 38, the baffle plates preferably overlap so as to provide a barrier of high resistance to flow. Both of these features can be accomplished by utilization of baffle plate extensions 44, as shown in FIG. 2a. The longitudinal clearances 46 accommodate the longitudinal expansion of the plates 34, and the transverse clearance 48 present a barrier to coolant flow. The transverse clearances 48 should therefore be sized as small as possible consistent with manufacturing techniques and maintenance of a generally smooth inner surface 40 and outer surface 42 throughout the entire height of the baffle assembly. For the embodiment shown, a transverse clearance 48 of 0.020 inches is consistent with these criteria. The longitudinal clearances 46 should be sized to accommodate, without significant interference, the expansions of consecutive plates 34. They will therefore vary dependent upon such parameters as the lengths of the baffle plates 34 and the temperatures the plates 34 are exposed to. In the embodiment shown in FIG. 2a, the upper longitudinal clearance is 0.120 inches, and the lower is 0.060 inches. The extensions 44 are approximately one inch long. FIG. 4 shows another embodiment which will maintain a zero transverse clearance 48. This is accomplished by providing a curved edge 50 on at least one of the extensions 44. As consecutive baffle plates 34 expand, the curved edge 50 will maintain contact with its mating extension 44, without presenting excessive resistance to the movement. FIGS. 5 through 7 present alternative embodiments which will also function to control thermal expansion and reduce fastener stresses without allowing excessive cross flow through the baffle plates at the junction regions 38. In FIGS. 5 and 5a consecutive baffle plates 34 are provided with male 52 and female 54 mating surfaces. The surfaces 52, 54 will allow longitudinal thermal expansions without interference, the expansions being taken within the longitudinal clearance 56. The male 52 and female 54 surfaces should therefore be sized to accept the expansion. The transverse clearances 58 should be sized to minimize the clearance area, without providing a significant resistance to the expansion movement. In this configuration, however, the clearances 56, 58 would allow cross flow through the junction region 38 unless otherwise prevented. Cross flow is therefore minimized by positioning the formers 36 to extend over the clearances. The formers 36 are preferably affixed to the baffles 34 by fasteners, such as bolts 60, through the male surface 52. FIG. 6 is an embodiment similar to that of FIG. 5a. Here, however, the transverse clearances 57 are enlarged to better facilitate attachment of the mating baffle plates 34 and increase some of the manufacturing tolerances. Here, the formers 36 are positioned to extend over the clearances. FIG. 7 shows another embodiment, similar to those of FIGS. 2a and 4, which also utilizes the formers 36 to minimize leakage across the baffle plates 34. The baffle plates 34 are provided with extensions 62 that can expand into the longitudinal clearances 64. The transverse clearance 66 is made as small as possible consistent with manufacturing techniques. Any leakage across the baffle plates 36 will therefore be minimized. To further alleviate leakage, the longitudinal clearances 64 on the baffle outer surface 42 are aligned with the formers 36. Aligning the longitudinal clearances 64 above or below the respective former 36 transverse centerline allows sufficient surface to affix the baffle 34 and former 36 by fastening means such as a bolt 68. Yet another embodiment is shown in FIG. 8. Here, a mating baffle plate is affixed to each former 36. Each baffle plate 34 is permitted to freely expand with temperature increases, without interfering with adjacent baffle plates 34. Thus, no significant thermally induced stresses will be imposed at the baffle-former attachment points 70. The baffle plates 34 have been shown with extensions 72, although flat edged plates can be utilized if flow is otherwise properly controlled to minimize core bypass flow and avoid vibration inducing impingement upon the fuel assemblies, such as those instances where any leakage flow would be from the core 14 outward. Affixing a separate baffle plate 34 to each former 36 provides the advantage of minimizing stresses, but would be more complex to manufacture and install than the other embodiments discussed. It is therefore seen that this invention provides a baffling arrangement which effectively baffles reactor coolant flow into and about the core of a nuclear reactor and which effectively reduces thermally induced stresses upon the baffling components. It will be apparent that many modifications and additions are possible in view of the above teachings. For example, mating baffle plates may be arranged at an angle, not necessarily presenting a horizontal upper or lower surface. Also, the number of plates may be varied, as may the geometric configuration and orientation of the mating extensions. It therefore is to be understood that within the scope of the appended claims, the invention may be practiced other than as specifically described.
046541721	description	DESCRIPTION OF THE PREFERRED EMBODIMENT Now, the fundamental principle of the present invention will be described. Methods of reducing the volume of a used ion exchange resin and converting it into inorganic matter include a wet process represented by acid decomposition and a dry process represented by a fluidized bed. The wet process involves the problem that the radioactive waste liquor containing a decomposition residue must be reprocessed by evaporation concentration or the like after the used resin is decomposed. The dry process is more advantageous than the wet process in that it is free from such aproblem, but the following problems occur in the fluidized bed process as a typical example of the dry process. (1 ) Large quantities of the residue and radioactive substances are spattered. In other words, since the used resin is decomposed and burnt under the fluidized gas, the residue and the radioactive substances are entrained and spattered by the exhaust gas. For this reason, the load to a filter for processing the exhaust gas becomes great. (2) Detrimental gases such as SOx or NOx are generated when the used resin is burnt, and the processing of the exhaust gas with an alkaline scrubber or the like becomes necessary, but the quantity of exhaust gas to be processed is enormous. In the fluidized bed process, air containing O.sub.2 3 to 5 times the chemical equivalent must be supplied and hence the exhaust gas quantity becomes great. (3) The radioactive waste after the volume reduction and inorganic conversion contains not only the residue but also Na.sub.2 SO.sub.4 and the like generated during the processing of the exhaust gas (SOx+NaOH.fwdarw.Na.sub.2 SO.sub.4 +H.sub.2 O). Accordingly, when 1 kg of the used resin is processed, the radioactive waste after the processing amounts to about 0.7 kg so that the volume reduction ratio is small. (4) Since the combustion is effected at a temperature of between 600.degree. and 900.degree. C., part of the residue is fused and deposited onto the furnace wall of the fluidized bed. If the fluidized bed is used for an extended period, the decomposition ratio will drop. (5) The non-fused radioactive residue that is withdrawn outside the furnace has a fine particle size (1 to 100 .mu.m), so that its handling is difficult. In order to solve these problems with the conventional fluidized bed process, the present invention provides a novel dry process which has the following constitutions to process the used ion exchange resin: (a) In order to prevent the residue and the radioactive substances from being spattered, the used ion exchange resin is pyrolyzed while it is kept in a stationary or like state. (b) The pyrolysis is effected at a low temperature (120.degree.-350.degree. C.) and then at a high temperature (350.degree. C. or above). (c) The residue after the pyrolysis is hot-pressed. Generally, an ion exchange resin is an aromatic organic high-molecular compound based on a copolymer of styrene and divinylbenzene (D.V.B.) and containing a sulfonic acid group bonded thereto in the case of a cation ion exchange resin and a quanternary ammonium group bonded thereto in the case of an anion exchange resin. In these resins, the bond energy between the ion exchange group (the sulfonic acid or quaternary ammonium group) and the resin main body is much weaker than that of the resin main body itself i.e. the copolymer between styrene and D.V.B. The present inventors have paid a special attention to this fact. When pyrolysis of the ion exchange resin is effected at a low temperature as a first-stage procedure, only the ion exchange group can be selectively decomposed. After the decomposition gas generated by this pyrolysis is separated, the remaining resin is pyrolyzed at a high temperature so as to decompose the resin main body and the resulting decomposition gas is separated. In this manner, nitrogen oxide gases (NOx) and sulfur oxide gases (SOx) that would otherwise need an elaborate exhaust gas treatment can be generated only in the first-stage low-temperature pyrolysis, while hydrogen gas (H.sub.2 ), carbon monoxide gas (CO) and carbon dioxide gas (CO.sub.2) that scarcely need the exhaust gas treatment can be generated selectively in the subsequent high-temperature pyrolysis. Accordingly, the quantity of the exhaust gases that must be processed can be drastically reduced, and the residue can be converted into stable inorganic compounds. In the present invention, when the ion exchange group is decomposed by the low-temperature pyrolysis, the feed of oxygen is not necessary so that the low-temperature pyrolysis can be effected in a stationary gas, thereby making it possible to prevent spattering of the residue and the radioactive waste. Since the secondary waste such as Na.sub.2 SO.sub.4 that is generated as a result of the exhaust gas treatment of NOx and SOx can be thus made nonradioactive, the radioactive waste is limited to only the residue after the high-temperature pyrolysis and the quantity of the radioactive waste after the pyrolysis can be drastically reduced to about 1/20. During the high-temperature pyrolysis for which the feed of oxygen is necessary, the velocity of the oxygen gas or air within the reaction vessel can be reduced to such an extent that the used resin does not spatter, and the spattering of the residue and the radioactive substances can be minimized. Thus, the load to a filter for treating the exhaust gas can also be reduced markedly. The present inventors have noted also the fact that the residue after the high-temperature pyrolysis is partly fused. Accordingly, in the present invention, this residue is hot-pressed into an easy-to-handle molded article, and the volume of the radioactive waste is reduced to about 1/30 of the original volume. Now, preferred embodiments of the present invention will be described in detail with reference to the accompanying drawings. The cation exchange resin has a cross-linked structure with a polymer backbone based on a copolymer consisting of styrene ##STR1## and divinylbenzene ##STR2## to which is bonded a sulfonic acid group (SO.sub.2 H) as the ion exchange group. It has a three-dimensional structure which is expressed by the following structural formula: ##STR3## Its molecular formula is (C.sub.16 H.sub.15 O.sub.3 S).sub.n. On the other hand, the anion exchange resin has a structure with a polymer backbone based on the same copolymer as that of the cation exchange resin, to which is bonded a quaternary ammonium group (NR.sub.3 OH) as the ion exchange group, and is expressed by the following structural formula: ##STR4## Its molecular formula is expressed by (C.sub.20 H.sub.26 ON).sub.n. Next, the bond energy at a bond portion between respective components of the ion exchange resin will be described. FIG. 1 shows the skeletal structure of the cation exchange resin, though that of the anion exchange resin is fundamentally the same except that the ion exchange group is different. The bond energy at each bond portion 1, 2, 3, and 4 between the respective components shown in FIG. 1 is listed in Table 1. TABLE 1 ______________________________________ Bond Bond energy portion Structure (kJ/mol) ______________________________________ 1 ion quaternary ammonium group 246 exchange (anion resin) group sulfonic acid group 260 (cation resin) 2, 3 polymer straight-chain portion 330-370 4 backbone benzene ring portion 480 ______________________________________ When the ion exchange resin is pyrolyzed, the ion exchange group having the smallest bond energy is first decomposed, then the straight-chain portion of the polymer backbone is decomposed and finally the benzene ring portion is decomposed. FIG. 2 shows the result of a thermogravimetric analysis (TGA) of the ion exchange resin using a differential thermal balance. However, the weight reduction resulting from the evaporation of water occurring at 70.degree. to 110.degree. C. is not illustrated. The solid line represents changes in the thermogravimetric weight of the anion exchange resin and the broken line that of the cation exchange resin. The decomposition temperature at each bond portion shown in FIG. 2 is listed Table 2. TABLE 2 ______________________________________ Decomposition temperature Structure (.degree.C.) ______________________________________ ion exchange group quaternary ammonium group 130-190 (anion resin) sulfonic acid group 200-300 (cation resin) polymer backbone straight-chain portion 350-400 benzene ring portion 380-480 ______________________________________ It can be understood from Table 2 that the quaternary ammonium group as the ion exchange group is first decomposed at 130.degree.-190.degree. C., then the straight-chain portion at 350.degree. C. or above and finally the benzene ring portion at 380.degree. C. or above in the anion exchange resin. In the cation exchange resin, on the other hand, the sulfonic acid group as the ion exchange group is first decomposed at 200.degree.-300.degree. C., then the straight-chain portion, and finally the benzene ring portion in the same way as in the anion exchange resin. In view of the result described above, only the ion exchange group of the ion exchange resin is first decomposed selectively in the first stage at a temperature of between 120.degree. and 350.degree. C., preferably about 300.degree. C., so that nitrogen and sulfur contained in only the ion exchange group are converted into nitrogen compounds (NOx, NH.sub.3, etc) and sulfur compounds (SOx, H.sub.2 S, etc) in this stage. Incidentally, a temperature of 120.degree. C. is a withstand temperature of the ion exchange resin, and the ion exchange group can be decomposed when being heated to at least this temperature. The temperature of 300.degree. C. is the point at which both the cation and anion exchange groups can be completely decomposed but the resin main body is not decomposed. Thereafter, the high-temperature pyrolysis is effected in the second stage at a temperature above 350.degree. C. Since the polymer backbone consisting of carbon and hydrogen is completely decomposed, the residue becomes below several percents. The exhaust gas generated at this time consists of CO, CO.sub.2, H.sub.2 and the like, so that no particular exhaust gas processing is necessary. Since the low-temperature and the high-temperature pyrolysis are carried out in multiple stages so as to decompose the ion exchange resin, the exhaust gas processing becomes by far easier than in the pyrolysis which is carried out in a single stage at a high temperature of above 350.degree. C. In other words, when the high-temperature pyrolysis is effected as the single stage treatment, 1.42 m.sup.3 of exhaust gas is generated per kg of ion exchange resin (a 2:1 mixture of an anion exchange resin and a cation exchange resin), and only about 5% of sulfur oxides and nitrogen oxides (0.074 m.sup.3 in total) are contained in the gas. If the pyrolysis is effected in two stages, on the other hand, the low-temperature pyrolysis is carried out below 350.degree. C. and then the high-temperature pyrolysis above 350.degree. C., so that 0.074 m.sup.3 of the sulfur oxides and nitrogen oxides are generated only in the low-temperature pyrolysis of the first stage, but they are not generated in the high-temperature pyrolysis of the second stage and 1.34 m.sup.3 of CO.sub.2 and the like is generated. Since emission of the exhaust gas into the air is legally regulated, the exhaust gas processing such as desulfurization or denitrification is necessary for the sulfur oxides and nitrogen oxides. Since they are generated only in a limited quantity during the low-temperature pyrolysis of the first stage, however, the quantity of the exhaust gas to be processed is only 0.074 m.sup.3. On the other hand, if the pyrolysis is effected as the single-stage treatment, great quantities of other exhaust gases must be altogether processed in order to process the sulfur and nitrogen oxides that are contained in a quantity of as small as only 0.074 m.sup.3 (5%), and the exhaust gases of as much as 1.42 m: must be processed. Accordingly, exhaust gas processing equipment must inevitably have a large scale. If the pyrolysis is carried out in two stages in accordance with the present invention, the quantity of the exhaust gases that must be carefully processed can be reduced to about 1/20. As described above, it has been found that if the ion exchange resin is pyrolyzed in two stages, the quantity of the exhaust gas that requires careful processing can be drastically reduced. In accordance with the fluidized bed process, the air containing oxygen two to five times the chemical equivalent must be supplied in order to fluidize the used resin, and hence the quantity of the exhaust gas that must be processed becomes enormous. In the present invention, on the other hand, the air to be supplied during the pyrolysis is extremely limited. This will be explained on the basis of the experimental results. The experimental results shown in FIGS. 3 and 4 pertain to the data when thermogravimetric analyses were carried out in an atmosphere of air containing oxygen in the chemical equivalent necessary for the pyrolysis of the used resin and in a nitrogen atmosphere not containing oxygen, respectively. Incidentally, the thermogravimetric analysis shown in FIG. 2 represents the data when oxygen in an amount sufficiently greater than the chemical equivalent was supplied. FIG. 3 represents the data when the cation exchange resin was pyrolyzed. The solid line represents the analysis effected in an atmosphere in which oxygen was present in the chemical equivalent, and the broken line represents the analysis effected in a nitrogen atmosphere. As shown in the diagram, the thermogravimetric characteristics similar to those when large quantities of oxygen was supplied could be observed if oxygen was present in an amount corresponding to the chemical equivalent, and the residue after the high-temperature pyrolysis could be reduced to below several percents. In the nitrogen atmosphere, too, the ion exchange group (sulfonic acid group) was pyrolyzed at 200.degree. to 300.degree. C. It was thus found that the feed of oxygen was not necessary for the pyrolysis of the ion exchange group. FIG. 4 shows the data when the anion exchange resin was pyrolyzed. In the same way as in FIG. 3, the solid line represents the atmosphere in which oxygen was present in an amount corresponding to the chemical equivalent, and the broken line represents the nitrogen atmosphere. It was found that in the pyrolysis of the anion exchange resin, too, the ion exchange group (quaternary ammonium group) could be decomposed at 130.degree. to 190.degree. C. even if no oxygen was present, and the polymer backbone could be decomposed at 350 to 480.degree. C. in the presence of oxygen in an amount corresponding to the chemical equivalent. It was found that oxygen need not be supplied in the low-temperature pyrolysis, and oxygen in an amount equal to, or greater than, the chemical equivalent need be supplied in the high-temperature pyrolysis. Thus, it can be understood that, in accordance with the present invention, the quantity of the exhaust gas to be processed can be reduced drastically. In accordance with the present invention, the spattering of the residue of the pyrolysis and the radioactive substances can be drastically reduced in comparison with the conventional fluidized bed process. Since the used resin is fluidized together with the gas in the fluidized bed process, the residue and the radioactive substances are entrained by the exhaust gas, resulting in the enhanced spattering. In accordance with the pyrolysis process, on the other hand, the spattering can be markedly reduced because the used resin can be calmly decomposed without causing its fluidization. This will be described with reference to FIGS. 5 through 7. FIG. 5 illustrates an apparatus used for the experiment. About 10 g of an ion exchange resin 6 containing about 100 .mu.Ci of adsorbed radioactive substances (.sup.58 Co, etc) was packed into a glass boat 5, and was thermally decomposed within a quartz tube 8. A tubular furnace 7 was used for the pyrolysis. Air 9 was supplied at a constant velocity from one of the ends of the quartz tube 8, and the quantities of the radioactive substances spattering towards the exhaust side and the amount of the residue were measured. FIG. 6 shows an example of changes in the spattering ratio of the radioactive substances when the pyrolysis temperature was changed. In the diagram, symbol C.P. and F.P. refer to a corrosive product and a nuclear fission product, respectively. The spattering ratio of .sup.58 Co represented by the solid line was below 10.sup.-3 % (detection limit) in the entire temperature range, while the spattering ratio of .sup.134 Cs represented by the broken line was below 10.sup.-3 % below 470.degree. C. and 0.2% above 470.degree. C. The spattering ratio of the residue was below 10.sup.-3 % in the entire temperature range for both .sup.58 Co and .sup.134 Cs. The reason why .sup.134 Cs spattered at a temperature above 470.degree. C. was that .sup.134 Cs adsorbed by the ion exchange group was oxidized by oxygen in the air into Cs.sub.2 O (m.p. 490.degree. C.) and this compound evaporated. To confirm this, the spattering ratios of other radioactive substances were also examined. As a result, it was found that the spattering started with temperatures above the melting points of their oxides. When the velocity of the air to be supplied into the quartz tube 8 was changed, the result shown in FIG. 7 could be obtained. In other words, the radioactive spattering ratio increased drastically at a velocity of above 1.5 cm/s, and it was in agreement with the spattering ratio of the residue. At a velocity below 1.5 cm/s, on the other hand, the spattering ratio of the residue was below 10.sup.-3 % in all cases, and the radioactive spattering ratio was also small. TABLE 3 ______________________________________ Melting point Radioactive spattering Radioactive of oxide initiating temperature nuclide (.degree.C.) (.degree.C.) ______________________________________ Corrosive .sup.58 Co 1800 >1000 product .sup.54 Mn 1650 (C.P.) .sup.59 Fe 1370 .sup.51 Cr 1550 Nuclear .sup.134 Cs 490 470 fission .sup.83 Rb 400 420 product .sup.90 Sr 2400 >1000 (F.P.) .sup.140 La 2000 ______________________________________ The results shown in FIG. 7 and Table 3 can be summarized as follows. (1) To reduce the quantities of the spattering residue and radioactive substances, the pyrolysis is preferably effected at a velocity of below 1.5 cm/s. (2) If the pyrolysis is effected at a velocity of below 1.5 cm/s, the spattering ratios of the radioactive substances are as follows: (i) below 10.sup.-3 % in all cases for corrosive products such as .sup.58 Co or .sup.54 Mn. (Generally, the radioactive substance contained in the used resin generated in a nuclear power station is only the corrosive product.) (ii) 10.sup.-3 % of nuclear fission products such as .sup.134 CS below 400.degree. C. and about 0.2% above 400.degree. C. (The nuclear fission products are contained in the used resin only when the breakage of fuel rods occurs.) When pyrolyzing the used resin, only the ion exchange group is selectively separated in the low-temperature pyrolysis (below 350.degree. C.) not requiring the feed of oxygen or the like, and the detrimental gas such as SOx is removed. Then, the polymer backbone is pyrolyzed in the high-temperature pyrolysis (above 350.degree. C.) while supplying oxygen in an amount at least equal to the chemical equivalent. In this manner, since no oxygen is supplied from outside during the low-temperature pyrolysis, the radioactivity of the exhaust gas such as SOx is extremely limited (radioactive spattering ratio <10.sup.-3 %), and the secondary waste generated as a result of the treatment of the exhaust gas such as SOx or NOx by an alkali scrubber or the like, such as Na.sub.2 SO.sub.4 (SOx+NaOH.fwdarw.Na.sub.2 SO.sub.4 +H.sub.2 O) and NaNO.sub.3 (NOx+NaOH.fwdarw.NaNO.sub.3 +H.sub.2 O), becomes non-radioactive. As a result, the radioactive waste is limited to only the residue. When 1 kg of the used resin (a 2:1 mixture of the cation exchange resin and the anion exchange resin) was processed, the radioactive spattering ratio was as high as from 10 to 20% in accordance with the conventional fluidized bed process, so that about 0.65 kg of the secondary waste such as Na.sub.2 SO.sub.4 and about 0.05 kg of the residue become the radioactive waste. In accordance with the present invention, on the other hand, only about 0.05 kg of the residue becomes the radioactive waste, so that the quantity of the radioactive waste can be drastically reduced. If the present invention is employed, the weight of the radioactive waste remaining after the inorganic conversion and volume reduction treatment of the used resin can be thus reduced to below 1/10 of the weight of the waste in accordance with the conventional fluidized bed process. Furthermore, since the velocity of the air supplied from outside during the high-temperature pyrolysis (above 350.degree. C.) is limited to below 1.5 cm/s in terms of the mean velocity within the reaction vessel in the present invention, the spattering of the residue as well as of the radioactive substances can be reduced remarkably (10.sup.-3 .about.0.2%). In comparison with the fluidized bed process in which the spattering ratio of the residue and radioactive substances is from 10 to 20%, the load to a filter for the exhaust gas can also be reduced remarkably. Incidentally, in the experiment shown in FIG. 5, a powdery ion exchange resin having an average particle size of 10 .mu.m was used as the ion exchange resin, though an about 20:1 mixture (volume ratio) of this resin and a granular ion exchange resin having an average particle size of 500 .mu.m is generally used in a nuclear power station. When only the granular ion exchange resin is processed, the spattering of the residue and radioactive substances does not occur if the average velocity of oxygen to be supplied is below 10 cm/s. In other words, in order to reduce the load to the filter for the exhaust gas, the air or oxygen must be supplied at such a level at which no spattering of the residue and radioactive substances will occur. As described above, if the used resin is pyrolyzed in two stages of the low-temperature and the high-temperature pyrolysis, the quantity of the exhaust gas that requires careful exhaust gas processing can be reduced to 1/20 and the weight of the radioactive waste can also be reduced to 1/10. Furthermore, the load to the filter for the exhaust gas can be reduced remarkably. The embodiments of the present invention, in which the two-stage pyrolysis method having the excellent features as described above is further developed, will now be described. Since the high-temperature pyrolysis is effected at 350.degree. C. or above, preferably from 500.degree. to 600.degree. C., part of the residue within the reaction vessel is in a fused state. For this reason, the residue sticks to the inner wall of the reaction vessel and cannot be easily withdrawn from the vessel. Accordingly, the reaction vessel can be used only 3 to ten times. The residue that can be withdrawn from the reaction vessel without sticking thereto is fine powder having a particle size of 1 to 100 .mu.m, and hence it is easy to spatter and its handling is not easy. The problem that part of the residue attaches to the reaction vessel is also observed in the conventional fluidized bed process, but in such a case, most of the residue is present in the fluidized gas so that the amount of deposition is as small as below 0.1% (5 to 10% in the two-stage pyrolysis method), and the vessel can be used repeatedly 50 to 200 times. (In the fluidized bed process, too, the heat transfer efficiency drops with the increase in the amount of deposition to thereby reduce the decomposition ratio of the used resin, and handling of the withdrawn residue is difficult, in the same way as in the two-stage pyrolysis method.) In order to solve the problems with the two-stage pyrolysis method described above, in the embodiments of the present invention, the residue is hot-pressed within the reaction vessel before it is withdrawn from the reaction vessel after the pyrolysis. One of such embodiments will be described in detail with reference to FIG. 8. The used resin 10 is placed in the reaction vessel 11 (FIG. 8(a)) and is then subjected to the volume reduction and inorganic conversion treatment (8(b)). The residue 12 generated in this case is hot-pressed as such while kept at the temperature of the high-temperature pyrolysis into a molded article 14 (8(c)). In this case, part of the residue 12 is in a fused state, so that it serves as a binder and a firm molded article 14 can be formed. Moreover, since the residue is at a high temperature, the pressure necessary for hot pressing is only about 1/10 of that effected at room temperature. Thereafter, the molded article 14 is withdrawn from the reaction vessel 11 (8(d) and 8(e)), and is stored in a waste storage vessel such as a drum 16 (8(f )). When hot-pressing the residue and withdrawing the molded article 14, upper and lower pistons 13 and 15 slide on the inner wall surface of the reaction vessel 11, so that any residue adherent to the inner wall surface of the reaction vessel can be completely removed, and build-up of the residue on the reaction vessel can be prevented. As an example, when 100 g of the used resin was packed in a cylindrical reaction vessel having an inner diameter of 40 mm and a depth of 200 mm and the resin was thermally decomposed at a high temperature of 600.degree. C., about 6 g of residue was left, and when this residue was hot-pressed at 600.degree. C. and a pressure of 50 kg/cm.sup.2 within the reaction vessel, there could be obtained a disc-like molded article having a volume of 6 cm.sup.3 and a density of 1 g/cm.sup.3. It was confirmed that the compression strength of this molded article became at least 150 kg/cm.sup.2 after cooling. For the sake of comparison, when about 6 g of residue was cooled and was then cold-pressed at a temperature of 20.degree. C. and a pressure of 500 kg/cm.sup.2 (the residue could not be cold-pressed at a pressure of 50 kg/cm.sup.2), there could be obtained a molded article having a density of 0.9 g/cm.sup.3, but its compression strength was as small as 10 kg/cm.sup.2. This suggests that, even if the two-stage pyrolysis is effected, the residue contains considerable organic matters and if the residue is hot-pressed under the high temperature condition where the residue is softened as a whole, molding can be effected under a pressure by far lower than that required for cold-press and moreover, part of the residue that is in a fused state functions as a binder in the case of hot-press, so that a molded article having by far higher strength can be obtained by hot-press than by cold-press. FIG. 9 shows the compression strength of the molded article after cooling when hot-press was effected under a pressure of 50 kg/cm.sup.2 while changing the hot-pressing temperatures. When hot-pressed at a temperature above 500.degree. C., the molded article exhibited a compression strength of at least 150 kg/cm.sup.2. When hot-pressed at a temperature below 350.degree. C., the molded article exhibited the compression strength below 100 kg/cm.sup.2. It was thus found that the strength of the molded article was low. Even when the apparatus of the invention described above was used repeatedly 100 times, no deposition nor build-up of the residue on the reaction vessel could be observed, and the drop of the decomposition ratio due to the use of the apparatus for an extended period could be prevented. As described above, when the residue after the two-stage pyrolysis is hot-pressed within the reaction vessel, the following effects can be obtained. (1) Deposition and build-up of the residue on the reaction vessel can be completely prevented and the apparatus can be used repeatedly more than 100 times. The heat transfer characteristics do not deteriorate during the use and the decomposition ratio of the used resin does not drop, either. (2) The molded article withdrawn from the reaction vessel is strong and does not get powdered. Accordingly, the residue can be handled extremely easily. (3) In accordance with the conventional fluidized bed process, the withdrawn residue is fine powder and is highly likely to spatter. Moreover, the bulk density of the residue is low (0.1-0.2 g/cm.sup.3). For this reason, the volume reduction effect is small and post-treatment such as pelletization or plastic solidification is necessary. In the embodiments of the present invention, on the other hand, the residue is hot-pressed under a pressure of about 50 kg/cm.sup.2 so that the molded article has a density of from 0.95 to 1.05 g/cm.sup.3. This value is extremely close to the true specific density of the residue of 1.1 g/cm.sup.3. Accordingly, the volume reduction effect is high and no post-treatment of the residue is necessary. In the embodiment described above, the hot-pressing temperature is the temperature of the high-temperature pyrolysis (ordinarily, from 500.degree. to 600.degree. C.), but hot-press may be effected at a higher temperature (about 800.degree. C.). In such a case, the proportion of the fused resin increases, so that the hot-pressing pressure can be reduced and the strength of the resulting molded article can be improved. The characterizing features of the embodiment described above can be summarized as follows. (1) The used resin is pyrolyzed in the two-stage pyrolysis consisting of the low-temperature and the high-temperature pyrolysis, and the residue after the pyrolysis is hot-pressed. (2) The pyrolysis and hot-press are carried out within the same reaction vessel. (3) In the low-temperature pyrolysis, the pyrolysis is conducted without feeding a gas such as oxygen at a temperature below 350.degree. C., while the high-temperature pyrolysis is conducted at above 350.degree. C. while feeding the air or oxygen gas. (4) Hot-press is effected in a stage in which part or the whole of the residue is fused or softened. Now, examples of the practical apparatus for embodying the method of the present invention described above will be described with reference to FIGS. 10 through 13. EXAMPLE 1 The apparatus shown in FIGS. 10 through 12 was used in the volume reduction and inorganic conversion of an ion exchange resin generated from a condensate purifier of a boiling water reactor by means of pyrolysis. FIG. 10 is a diagram showing the construction of the system, FIG. 11 is a prespective view of part of the reaction apparatus, and FIG. 12 is a schematic sectional view of the apparatus. The waste resin took a slurry form because it was discarded from a condensate desalting device by back wash. The waste resin slurry containing corrosive products such as .sup.60 Co or .sup.54 Mn as the radioactive substances was supplied from a slurry transportation pipe 17 to a slurry tank 18. A predetermined quantity of the waste resin within the slurry tank 18 was supplied to a reaction vessel 40 provided in the reaction apparatus 24 through a valve 22. A plurality (ten in this example) of reaction vessels 40 were disposed on a turn table 38 in the disc arrangement as shown in FIG. 11, and each reaction vessel had an inner volume of 300 l and a diameter of 550 mm.phi.. The waste resin containing adsorbed corrosive products such as .sup.60 Co in an amount of 10.sup.-2 .mu.Ci/g (on a dry basis) was supplied to each reaction vessel 40 in an amount of 10 kg (100 kg in total). After the resin was supplied, a lid 52 leading to an exhaust gas processing system was placed, and the waste resin supplied into each reaction vessel 40 was heated to 350.degree. C. by a heater 34 for pyrolysis without feeding oxygen or the like as an oxidizing agent. As a result, only the ion exchange group of the waste resin was pyrolyzed, producing about 10 m.sup.3 of sulfur and nitrogen compounds (SOs, H.sub.2 S, NOx, NH.sub.3, etc) in the gas form. These gases were introduced into the exhaust gas processing apparatus through the valve 23, were removed in an alkali scrubber 31 by an aqueous solution of sodium hydroxide supplied from a feed pipe 29, and were converted into an aqueous solution of sodium salts (Na.sub.2 SO.sub.4, NaNO.sub.3, etc). The solution was discharged through a discharge pipe 30. Since these aqueous solutions are non-radioactive, they can be processed by non-radioactive chemical waste liquor processing steps in the nuclear power station. When the waste liquor obtained in this example was dried, the radioactive concentration of the resulting solid Na.sub.2 SO.sub.4 and the like was below 10.sup.-7 .mu.Ci/g, which is the detection limit by a current precision measurement method, and the secondary waste such as Na.sub.2 SO.sub.4 could be handled as the non-radioactive waste. This means also that the contamination removal coefficient in the low-temperature pyrolysis is at least 10.sup.5. Incidentally, the moisture contained in the waste resin was generated as vapor, and the vapor was condensed by a condenser 27 and was recovered as the water for re-use from the pipe 28. A considerable amount of exhaust gas after the treatment by the alkali scrubber 31 was discharged through a filter 32. After the low-temperature pyrolysis was made in the course of about one hour in the reaction vessels 40, the remaining waste residue (consisting solely of the polymer backbone) was pyrolyzer at a high temperature of 600.degree. C. by the heater 34 in the same vessels 40. During this high-temperature pyrolysis, the air from an air pump 19 was continuously supplied into each reaction vessel 40 at a rate of 150 l/min through the valve 21. As a result, the average velocity in the reaction vessel became about 1 cm/sec. After the high-temperature pyrolysis for about 6 hours, the polymer backbone could be decomposed, and only the stable residue was left in an amount of about 0.5 kg in each reaction vessel 40. About 200 m.sup.3 of carbon dioxide (CO.sub.2), carbon monoxide (CO), hydrogen gas (H.sub.2), hydrocarbon gas (CH.sub.4) and the like were generated by the high-temperature pyrolysis, and these exhaust gases passed through the valve 35 and the filter 25 for the high-temperature decomposition, then entered a flare stack 26, whereby they were burnt and exhausted as CO.sub.2 and H.sub.2 O gases. The quantity of the radioactive substances contained in the exhaust gases and collected by the filter 25 was measured, but the radioactivity was below the detection limit. The contamination removal coefficient in the high-temperature pyrolysis was at least 10.sup.4. The quantity of the residue collected by the filter 25 was below 5 g, and the load to the filter was reduced extremely. The residue after the high-temperature pyrolysis was hot-pressed by upper and lower presses 43 and 47 at a pressure of 40 kg/cm.sup.2 (total pressure: 100 ton) while it was kept at the high-temperature pyrolysis point of 600.degree. C. in the same reaction vessels 40. After the hot-press, the residue was turned into a disc-like molded article 50, moved downwards together with the piston 48a of the hydraulic cylinder 48 of the lower press 47, was discharged by the hydraulic cylinder 46, was charged in a drum 49 and was finally solidified by a solidifying agent such as cement or plastics. The undecomposed polymer backbone of the waste resin was decomposed by the high-temperature pyrolysis to be converted into a stable inorganic residue. Accordingly, it was extremely stable to store. The residue after the decomposition consisted primarily of silica (SiO.sub.2) and a clad (mainly iron oxides) in the cooling water for the reactor, that attached to the ion exchange resin. After the hot-pressing of the residue in the reaction vessel 40 was completed, the turn tables 38, 39 were rotated by 1/10 with a shaft 41 being the center, and the adjacent reaction vessel 40 containing only the residue after the high-temperature pyrolysis was moved to the position of the presses 43, 47 so that the residue was hot-pressed in the same way as described above. In this manner, the waste resin charged in the reaction vessel 40 was subjected to the two-stage pyrolysis, the remaining residue was sequentially hot-pressed and was sequentially charged in the drum. Though part of the residue attached to the inner wall surface of the reaction vessel 40, the remaining residue was scraped off by the pistons 44a, 48a of the cylinders 44, 48 when the upper and lower cylinders 44, 48 slid inside the reaction vessel 40. Thus, all the residue could be converted into the molded article. In accordance with this example, both low and high temperature pyrolysis and hot-press could be carried out in the same reaction vessel 40, and the volume reduction and inorganic conversion of the waste resin could be efficiently effected without permitting any residue to remain in the reaction vessel 40. Since the resulting molded article 50 had a sufficiently high strength, it could be easily handled without undergoing powdering or breakage. Furthermore, the molded article 50 had a density of as great as 0.9 g/cm.sup.3 and exhibited a high volume reduction effect. In other words, when 100 kg of the used resin was processed, the resulting radioactive waste was only 5 kg of the residue, and its volume was about 5.5 l (about 1/30 of the original volume). Accordingly, the volume of the radioactive waste dropped below 1/5 in comparison with the conventional fluidized bed process and acid decomposition process. In this example, the air was supplied as the oxidizing agent for the high-temperature pyrolysis, but oxygen can be also supplied. In such a case, if oxygen is supplied at the same feed speed as that of the air, the time necessary for the high-temperature pyrolysis can be reduced by maximum 1/5, but the possibility of explosion is induced. FIG. 13 illustrates the effect of the addition an oxidizing agent. In the drawing, in the case of the nitrogen atmosphere without the addition of the oxidizing agent in the high-temperature pyrolysis of 350.degree. C. or above (represented by curve A), about 25 to 30% of residue remained even if heating was made to 1,000.degree. C. On the other hand, when steam was added as the oxidizing agent (represented by curve B), the residue could be drastically reduced at 600.degree. C. or above, and dropped below several percents at 700.degree. C. or above. When the air was used as the oxidizing agent (represented by curve C), the weight dropped drastically at 400.degree. C. or above, and the residue dropped below several percents at 500.degree. C. or above. In other words, the high-temperature pyrolysis in the reaction vessel 40 is preferably carried out at a temperature of above 700.degree. C. if the inert gas such as nitrogen gas is used, and at a temperature of above 500.degree. C. if the pyrolysis is made in an atmosphere of air. In order to minimize the residue, the oxidizing agent such as steam or air is preferably added. This makes it possible to reduce the volume of the waste resin to about 1/10. In the example described above, the low and high temperature pyrolysis as well as hot-pressing were effected in the same reaction vessel, but they can be, practiced in separate vessels. In such a case, the operation procedures become more complicated. The vessel in which hot-pressing is made must be sufficiently strong to withstand the pressure. The example described above is related to an application to the boiling water reactor, but the present invention can also be applied to the processing of the used ion exchange resin generated in waste liquor purification systems of installations handling the radioactive substances, such as a reactor purification system, a primary coolant purification system of a pressurized water reactor, and so forth. In the example described above, the exhaust gas generated during the low-temperature pyrolysis was processed by use of the alkali scrubber 31, but the same effect can be obtained by dry processing of the exhaust gas using active carbon, MnO, or the like. In the example described above, the temperatures in the low and high temperature pyrolysis were controlled by the heater 34, the thermometer 36 and the controller 37, and the operation of the valves 23 and 35 for the two exhaust gas systems were also controlled by the controller 37. Before pyrolyzing the ion exchange resin, the moisture contained in the resin may be removed by heating or centrifugal means before the resin is charged in the reaction vessel 40 or by heating the resin to 110.degree. to 120.degree. C. by the heater 34 after the resin is charged in the reaction vessel 40. EXAMPLE 2 Example 1 pertains to the example of the volume reduction and inorganic conversion of the used ion exchange resin containing only the adsorbed corrosive products (Co, Mn, Fe, etc) as the radioactive substances. An experiment of processing a used ion exchange resin containing adsorbed nuclear fission products (Cs, Sr, etc) was carried out to cope with the possibility of breakage of nuclear fuel rods. 100 kg of used ion exchange resins containing 10.sup.-2 .mu.Ci/g (dry weight) of the adsorbed corrosive products and the nuclear fission products, respectively, were processed in the same way as in Example 1. As a result, exactly the same result could be obtained as in Example 1 except the following point. The difference was that among the nuclear fission products generated by the high-temperature pyrolysis, the radioactive substances whose oxides had a low melting point, such as Cs and Rb, spattered and were collected by the filter 25 for the high-temperature pyrolysis. For this reason, the contamination removal coefficient in the high-temperature pyrolysis became about 10.sup.3, but the load to the filter was by far smaller than that in the conventional fluidized bed process (contamination coefficient: 10.about.20). EXAMPLE 3 The low-temperature pyrolysis was effected at 350.degree. C. in Example 1, but it can be carried out at a temperature equal to the high-temperature pyrolysis, for example, at 600.degree. C. As can be seen clearly from FIGS. 3 or 4, only the ion exchange group can be decomposed and removed even if pyrolysis is effected at a temperature of 350.degree. C. or above without feeding oxygen. For example, pyrolysis was first made at 600.degree. C. without feeding oxygen to remove the ion exchange group, and the polymer backbone was then pyrolyzed at the same temperature of 600.degree. C. by feeding oxygen. In such a case, the apparatus could be simplified, but if the used resin had adsorbed those radioactive substances which were easily spattered, such as Cs and Rb, these radioactive substances would be incorporated in the secondary waste such as Na.sub.2 SO.sub.4 that was generated as a result of the exhaust gas processing of sulfur and nitrogen compounds (SOx, H.sub.2 S, NOx, NH.sub.3, etc), so that the amount of the radioactive waste became about 5 times that of Example 1. Accordingly, this example exhibited a remarkable effect in processing the used resin which had adsorbed only the corrosive products such as Co or Mn. EXAMPLE 4 Only the residue was hot-pressed in Example 1, but it is also effective to charge in advance a vitrifying agent corresponding to 10 to 40 wt % of the residue generated finally, and then to carry out hot-pressing after the resin is pyrolyzed in two stages. In other words, the vitrifying agent is in a fused state during the hot-pressing so that it functions as a binder and the pressure necessary for the hot-pressing needs be only about 1/2 of the pressure (40 kg/cm.sup.2) in Example 1. In addition, when the molded article 50 is finally solidified in the waste storage vessel such as a drum 49, the vitrifying agent has high affinity with the molded article and with the solidifying agent, so that the durability of the solidified waste can be improved. The radioactive substances that are easily spattered, such as Cs and Rb, are entrapped in the network structure of the glass during the high-temperature pyrolysis and are solidified and fixed. For this reason, the radioactive spattering ratio can be improved extremely remarkably. An ordinary glass frit consisting principally of silica (SiO.sub.2) may be used as the vitrifying agent. Since the glass frit is fused at 500.degree. to 600.degree. C., it functions as the binder and also entraps Cs, thus preventing spattering of Cs. It is also preferred to add about 20 wt % of boron oxide (B.sub.2 O.sub.3) during the pyrolytic reaction in order to carry out efficiently the fusing and solidification of the glass. In this case, the vitrifying agent acts effectively only during the high-temperature pyrolysis, but from the viewpoint of the operation procedures, the vitrifying agent is preferably charged in the reaction vessel 40 together with the waste resin before carrying out the low-temperature pyrolysis. In FIG. 10, reference numeral 33 represents a glass frit feed pipe, and an arbitrary amount of the glass frit is fed to the reaction vessel 40 by the operation of the valve 20. EFFECTS OF THE INVENTION In the present invention, the used ion exchange resin is pyrolyzed by the two-stage pyrolysis at low and high temperatures, and the resulting residue is hot-pressed. Accordingly, the present invention can drastically reduce the volume, and can selectively process the exhaust gases generated during the pyrolysis.
	claims	1. A loaded polymer sheet loaded with a high atomic weight metal, and useful for forming a protective garment, wherein the sheet is prepared from a polymer latex liquid having dispersed therein a high atomic weight metal having an atomic number greater that 45, wherein the quantity of the loaded high atomic weight metal in the polymer sheet exceeds 89 percent by weight of the total loaded polymer sheet, including the polymer and the metal, and wherein the thickness of the loaded sheets required to achieve the radiation attenuation equivalent to 0.5 mm of a pure lead sheet has a weight of less than about 1.0 pound/ square foot. 2. The loaded polymer sheet of claim 1 wherein the metal is selected from the group consisting of antimony, tin, bismuth, tungsten, lead, cadmium, indium, cesium, cerium and gadolinium and any combination thereof. 3. The loaded polymer sheet of claim 1 having a thickness in the range of from about 0.010 inches to about 0.05 inches and not greater than about 0.05 mm. 4. The loaded polymer sheet of claim 1 wherein the polymer is selected from the group consisting of natural and synthetic polymers. 5. The loaded polymer sheet of claim 4 wherein the polymer is selected from the group consisting of acrylic, styrene/butadiene, vinyl acetate/acrylic acid copolymers, vinyl acetate, ethylene vinyl acetate, polybutene, and urethane polymers, and natural rubber and combinations thereof. 6. The loaded polymer sheet of claim 1 wherein the polymer sheet is formed from a fluid polymer latex having a pH value of above 8.5 and with at least one high atomic weight metal in particulate form dispersed therein in an amount of at least 89% by wt. of the combined polymer and metal particles, the latex being sufficiently fluid to be able to be poured to cast a sheet on a flat substrate. 7. The loaded polymer sheet of claim 6 wherein the metal particles having an average particle size of at least about 8 microns. 8. The loaded polymer sheet of claim 7 wherein the polymer is an elastomer and the metal particles have an average particle size of at least about 10 microns. 9. The method of producing a loaded polymer sheet comprising the steps of: mixing a high atomic weight metal in particulate form into a polymer latex having a pH of at least 8.5, wherein the high atomic weight metal has an atomic number greater then 45, and exceeds about 89 percent by weight of the total polymer plus metal in the latex, casting the latex on a flat surface, and drying the cast latex to form a useful loaded polymer sheet that weighs less than about 1.0 pound/square foot at a thickness sufficient to achieve the equivalent radiation attenuation as a pure lead sheet having a thickness of 0.5 mm. 10. The method of claim 9 wherein the metal is selected from the group consisting of antimony, tin, bismuth, tungsten, lead, and any combination thereof. 11. The method of claim 9 wherein the metal is selected from the group consisting of cadmium, indium, cesium, cerium and gadolinium and any combination thereof. 12. The method of claim 9 wherein the thickness of the sheet is at least about 0.010 inch. 13. The method of claim 12 wherein the thickness of the sheet is in the range of from about 0.015 inch to about 0.07 inch. 14. The method of claim 9 wherein the polymer latex is selected from the group consisting of natural and synthetic polymers. 15. The method of claim 14 wherein the polymer latex is selected from the group consisting of acrylic polymers, styrene/butadiene copolymers, vinyl acetate/acrylic acid copolymers, vinyl acetate polymers, ethylene vinyl acetate polymers, polybutene polymers, urethane polymers and combinations thereof 16. The method of claim 14 wherein an additive selected from the group consisting of surfactants, defoamers, antifoaming agents, dispersing aids and plasticizers is incorporated into the latex. 17. The method of claim 14 wherein the polymer latex is selected from the group of mixed polymers consisting of ethylene vinyl acetate and acrylic coplymers, acrylic and styrene acrylic polymers, polybutene and natural rubber polymers, polybutene and acrylic polymers, styrene-butadiene and styrene acrylic polymers, and isoprene and acrylic polymers. 18. The method of claim 9 comprising the additional step of: after the mixture is dried, applying a coating of unfilled latex to a surface of the dried loaded polymer sheet. 19. The method of claim 18 wherein a thickness of the coating is in the range of about 0.25 mils to about 4 mils. 20. The method of producing a loaded polymer sheet comprising the steps of: mixing particulate tungsten metal into a polymer latex; adding particulate tin to the mixture, such that the total amount of the combination of tin and tungsten exceeds about 89 percent by weight of the total weight of polymer and metal; and drying the mixture to form a loaded polymer sheet that weighs less than about 1.0 pound/square foot at a thickness of loaded polymer sheet required to achieve the equivalent radiation attenuation as 0.5 mm thickness of a pure lead sheet. 21. The method of claim 20 wherein the polymer latex comprises a natural rubber latex. 22. A polymer latex, comprising dispersed polymer and a high atomic weight metal in particulate form, wherein the amount of the high atomic weight metal exceeds about 89 percent by weight of the total polymer plus metal in the latex, the latex having a pH of at least about 8.5 and a viscosity sufficiently low to permit casting the latex on a flat surface. 23. The loaded polymer sheet of claim 3 having a thickness of in the range of from about 0.015 inches to about 0.05 inches. 24. The loaded polymer sheet of claim 1 wherein the metal is selected from the group consisting of antimony, tin, bismuth, tungsten, and any combination thereof. 25. The loaded polymer sheet of claim 1 wherein the metal is selected from the group consisting of cadmium, indium, cesium, cerium and gadolinium and any combination thereof. 26. The method of claim 9 wherein the metal is selected from the group consisting of antimony, tin, bismuth, tungsten, cadmium, indium, cesium, cerium and gadolinium and any combination thereof. 27. The polymer latex of claim 22, wherein the metal is selected from the group consisting of antimony, tin, bismuth, tungsten, cadmium, indium, cesium, cerium and gadolinium and any combination thereof. 28. The loaded polymer sheet of claim 1, wherein the sheet is flexible and where the metal is selected from the group consisting of antimony, tin, bismuth, tungsten, cadmium, indium, cesium, cerium and gadolinium and any combination thereof.
	summary
059129360	claims	1. A pipe connector assembly for coupling a first pipe section to a second pipe section in a nuclear reactor, said pipe connector assembly comprising: a first coupling member comprising a flange, a substantially cylindrical pipe engaging portion extending from a first surface of said flange and a spherical convex seat portion extending from a second surface of said flange; and a second coupling member comprising a flange having a spherical concave seat portion for receiving said convex seat portion of said first coupling member and a substantially cylindrical pipe engaging portion extending from a first surface of said second coupling member flange, said first coupling member flange comprising at least one stud bore extending therethrough and said second coupling member flange comprising at least one stud bore extending therethrough, said stud bores configured to be elongate in an azimuthal direction. welding the first pipe section to the first coupling member; welding the second pipe section to the second coupling member; and coupling the first coupling member to the second coupling member. a first pipe section; a second pipe section; and a first pipe connector assembly coupling said first pipe section to said second pipe section, said first pipe connector assembly comprising a first coupling member comprising a flange, a substantially cylindrical pipe engaging portion extending from a flange first surface, and a convex spherical seat portion extending from a flange second surface, said first pipe connector assembly further comprising a second coupling member comprising a flange having a concave spherical seat portion for receiving said convex spherical seat portion of said first coupling member and a substantially cylindrical pipe engaging portion extending from a first surface of said second coupling member, said first coupling member pipe engaging portion configured to engage said first pipe section, said second coupling member pipe engaging portion configured to engage said second pipe section, said second coupling member concave spherical seat portion configured to seat on said first coupling member convex spherical seat portion, said first coupling member flange further comprising at least one stud bore extending therethrough, and said second coupling member flange further comprising at least one stud bore extending therethrough, said stud bores configured to be elongate in an azimuthal direction. a third pipe section; a second pipe connector assembly coupling said second pipe section to said third pipe section, said second pipe connector assembly comprising a first coupling member comprising a flange, a substantially cylindrical pipe engaging portion extending from a flange first surface, and a convex spherical seat portion extending from a flange second surface, said second pipe connector assembly further comprising a second coupling member comprising a flange having a concave spherical seat portion for receiving said convex spherical seat portion of said first coupling member and a substantially cylindrical pipe engaging portion extending from a first surface of said second coupling member, said first coupling member pipe engaging portion configured to engage said second pipe section, said second coupling member pipe engaging portion configured to engage said third pipe section, said second coupling member concave spherical seat portion configured to seat on said first coupling member convex spherical seat portion. 2. A pipe connector assembly in accordance with claim 1 wherein said first coupling member pipe engaging portion is configured to engage the first pipe section and said second coupling member pipe engaging portion is configured to engage the second pipe section. 3. A pipe connector assembly in accordance with claim 1 wherein at least one locking element is configured to extend through said first and second coupling member stud bores. 4. A pipe connector assembly in accordance with claim 3 wherein said stud bores have a diameter larger than a diameter of said locking element. 5. A pipe connector assembly in accordance with claim 3 wherein said locking element further comprises at least one spherical washer. 6. A pipe connector assembly in accordance with claim 3 wherein said locking element further comprises at least one crimp locking mechanism. 7. A pipe connector assembly in accordance with claim 1 wherein said first coupling member convex spherical seat portion is configured to seat portion on said second coupling member concave spherical seat portion when said first coupling member is coupled to said second coupling member. 8. A pipe connector assembly in accordance with claim 7 wherein said first coupling member spherical seat portion is rotatable relative to said second coupling member spherical seat portion when said first coupling member is coupled to said second coupling member. 9. A pipe connector assembly in accordance with claim 7 wherein said convex and concave spherical seat portions allow up to four degrees of rotational misalignment when said first coupling member is coupled to said second coupling member. 10. A method of coupling a first pipe section to a second pipe section utilizing a pipe connector assembly, said pipe connector assembly having a first coupling member comprising a flange, a substantially cylindrical pipe engaging portion extending from a first surface of the flange and a spherical convex seat portion extending from a second surface of the flange, and a second coupling member comprising a flange having a spherical concave seat portion for receiving the convex seat portion of the first coupling member and a substantially cylindrical pipe engaging portion extending from a first surface of the second coupling member flange, the first coupling member flange comprising at least one stud bore extending therethrough and the second coupling member flange comprising at least one stud bore extending therethrough, the stud bores configured to be elongate in an azimuthal direction, said method comprising: 11. A method in accordance with claim 10 wherein said method further comprises the steps of welding the first pipe section to the first coupling member pipe engaging portion, and welding the second pipe section to the second coupling member pipe engaging portion. 12. A method in accordance with claim 10 wherein said method further comprises the step of seating the first coupling member convex seat portion on the second coupling member concave seat portion. 13. A method in accordance with claim 12 wherein said method further comprises the step of securing at least one locking element through the stud bores to couple the first and second coupling members. 14. A method in accordance with claim 13 further comprising the step of coupling at least one spherical washer and a crimp locking mechanism to the locking element. 15. A replacement core spray line, comprising: 16. A replacement core spray line in accordance with claim 15 further comprising: 17. A replacement core spray line in accordance with claim 16 further comprising a third pipe connector assembly coupling said third pipe section to a nozzle junction, said third pipe connector assembly comprising a first coupling member comprising a flange, a substantially cylindrical pipe engaging portion extending from a flange first surface, and a convex spherical seat portion extending from a flange second surface, said third pipe connector assembly further comprising a second coupling member comprising a flange having a concave spherical seat portion for receiving said convex spherical seat portion of said first coupling member and a substantially cylindrical pipe engaging portion extending from a first surface of said second coupling member, said first coupling member pipe engaging portion configured to engage said third pipe section, said second coupling member pipe engaging portion configured to engage said nozzle junction, said second coupling member concave spherical seat portion configured to seat on said first coupling member convex spherical seat portion, and wherein said first pipe section is coupled to a shroud connector forming a fluid passage between the shroud connector and the nozzle junction. 18. A replacement core spray line in accordance with claim 15 wherein at least one locking element is configured to extend through said stud bores to couple said first coupling member to said second coupling member. 19. A replacement core spray line in accordance with claim 18 wherein said locking element further comprises at least one of a spherical washer and a crimp locking mechanism.
046735445	abstract	A pushing device for removing spent fuel rods from a nuclear reactor fuel assembly comprises a plurality of axially shiftable rods disposed in the same geometric pattern and with the same pitch as the fuel rods. The push rods are mounted at their upper ends to a pressure strip or plate and traverse at their lower ends bores in a guide plate fixed via guide rods to a base plate located on the other side of the pressure plate, the pressure plate being movably secured to the guide rods for linear motion between the base plate and the guide plate. A drive mounted to the base plate is connected to the pressure plate for shifting the latter, while a safety mechanism is provided for interrupting the motion of a push rod if the respective fuel rod is jammed in the fuel assembly.
042773067	claims	1. In a toroidal plasma confinement method including the steps of providing a toroidal vacuum chamber zone, providing a high temperature hydrogen plasma in the vacuum chamber zone and a toroidal magnetic field in which the plasma is embedded, and providing a poloidal magnetic field in which the plasma is embedded, the improvement comprising controlling plasma impurities by generating a magnetic plasma confining field for confining a high temperature plasma having continuous inner flux lines in a high temperature plasma zone in a plasma vacuum chamber without internal divertor coils, which chamber is elongated in a direction along the major toroidal axis, generating a region of lower magnetic field strength in said elongated vacuum chamber for confining a plasma of lower temperature than said high temperature plasma to provide said poloidal field with a field configuration in which the radially outer flux lines of said poloidal magnetic field about the high temperature plasma are axisymmetrically bound to terminate at the vacuum chamber wall about said region of lower magnetic field strength adjacent the high temperature plasma in a direction along the major toroidal axis opposite the direction of positive ion toroidal drift, providing a source of pure hydrogen at the high temperature plasma zone, evacuating the low temperature plasma zone, and maintaining a low temperature plasma adjacent the high temperature plasma. 2. A method in accordance with claim 1 wherein said low temperature plasma in maintained by r-f heating. 3. A method in accordance with claim 1 wherein impurity control is accomplished at least in part by collisions between protons and impurity ions sufficient to produce collisional pumping. 4. A method in accordance with claim 1 wherein said impurity control is accomplished at least in part by electrostatic pumping.
	abstract	A heat dissipation structure includes a housing. The housing has a bottom surface, a liquid inlet channel, a liquid outlet channel and a protruding portion. The liquid inlet channel and the liquid outlet channel are located at two opposite ends of the housing and above the bottom surface. The liquid inlet channel and the liquid outlet channel extend along a first direction. The protruding portion is located between the liquid inlet channel and the liquid outlet channel and above the bottom surface. The protruding portion protrudes towards a direction away from the bottom surface. The protruding portion has a protruding surface facing away from the bottom surface. A distance between the protruding surface and the bottom surface is increased first and then decreased along the first direction.
043022847	claims	1. A toroidal plasma device comprising a toroidal confinement vessel having walls for defining a toroidal space and confining gas therein, means for generating magnetic flux linking said toroidal space to induce substantial toroidal plasma current therein, said toroidal plasma current producing a substantial poloidal magnetic field, first and second windings wound substantially helically around said vessel with the same sense of twist at substantially the same pitch, said first windings and said second windings being disposed alternately and substantially equally spaced around the minor circumference of said vessel, and means for passing first direct current through said first windings and second direct current through said second windings in the direction counter to said first direct current to generate a helical magnetic field acting in combination with said poloidal magnetic field to produce closed and nested magnetic flux surfaces spaced from said vessel walls, wherein a safety factor q within said plasma current is the sum of two components, one being axisymmetric and substantially proportional to the ratio of toroidal magnetic field to poloidal magnetic field, and the other being nonaxisymmetric and substantially helically symmetric and substantially the quantity ##EQU41## where b is a measure of the strength of the magnetic field from said helical windings, R is the major radius of said toroidal space, r is the average minor radius of the magnetic flux surface, B.sub..theta.,o is the poloidal magnetic field produced by said toroidal plasma current, ##EQU42## is the partial derivative of B.sub..theta.,o with respect to r, l is the number of said first windings, k is the wavenumber of the magnetic field produced by said first and second windings, I.sub.l (kr) is the modified Bessel function of order l, and I.sub.l '(kr) is the derivative of I.sub.l (kr) with respect to its argument, q being defined as the average over a flux surface of the number of transits made around said toroidal space in the toroidal direction by a magnetic flux line in making a single transit in the poloidal direction, and the absolute magnitude of q being less than 1 within said plasma current. a toroidal confinement vessel for defining a toroidal space and confining gas therein, means for generating magnetic flux linking said toroidal space to induce substantial toroidal plasma current therein, said toroidal plasma current producing a substantial poloidal magnetic field, first and second windings wound substantially helically around said vessel with the same sense of twist at substantially the same pitch, said first windings and said second windings being disposed alternately and substantially equally spaced around the minor circumference of said vessel, and means for passing first direct current through said first windings and passing second direct current through said second windings in the direction counter to said first direct current to generate a helical magnetic field acting in combination with said poloidal magnetic field to produce a variation in a safety factor q with minor radius at any poloidal angle whereby the polarity of q reverses near the outer edge of the plasma current, q being defined as the average over a flux surface of the number of transits made around said toroidal space in the toroidal direction by a magnetic flux line in making a single transit in the poloidal direction, wherein a safety factor q within said plasma current is the sum of two components, one being axisymmetric and substantially proportional to the ratio of toroidal magnetic field to poloidal magnetic field, and the other being nonaxisymmetric and substantially helically symmetric and substantially the quantity ##EQU43## where b is a measure of the strength of the magnetic field from said helical windings, R is the major radius of said toroidal space, r is the average minor radius of the magnetic flux surface, B.sub..theta.,o is the poloidal magnetic field produced by said toroidal plasma current, ##EQU44## is the partial derivative of B.sub..theta.,o with respect to r,l is the number of said first windings, k is the wavenumber of the magnetic field produced by said first and second windings, I.sub.l (kr) is the modified Bessel function of order l, and I.sub.l '(kr) is the derivative of I.sub.l (kr) with respect to its argument, and q being less than 1 within said plasma current. a toroidal confinement vessel having walls for defining a toroidal space and confining gas therein, means for generating magnetic flux linking said toroidal space to induce substantial toroidal plasma current therein, said toroidal plasma current producing a substantial poloidal magnetic field, means for generating a vertical magnetic field within said toroidal space, said vertical field being perpendicular to the equatorial plane of said toroidal space, first and second windings wound substantially helically around said vessel with the same sense of twist at substantially the same pitch, said first windings and said second windings being disposed alternately and substantially equally spaced around the minor circumference of said vessel, and means for passing first direct current through said first windings and second direct current through said second windings in the direction counter to said first direct current to generate a helical magnetic field acting in combination with said poloidal magnetic field and said vertical field to produce closed and nested magnetic flux surfaces spaced from said vessel walls, wherein a safety factor q within said plasma current is the sum of two components, one being axisymmetric and substantially proportional to the ratio of toroidal magnetic field to poloidal magnetic field, and the other being nonaxisymmetric and substantially helically symmetric and substantially the quantity ##EQU45## where b is a measure of the strength of the magnetic field from said helical windings, R is the major radius of said toroidal space, r is the average minor radius of the magnetic flux surface, B.sub..theta.,o is the poloidal magnetic field produced by said toroidal plasma current, ##EQU46## is the partial derivative of B.sub..theta.,o with respect to r, l is the number of said first windings, k is the wavenumber of the magnetic field produced by said first and second windings, I.sub.l (kr) is the modified Bessel function of order l, and I.sub.l '(kr) is the derivative of Il(kr) with respect to its argument, q being defined as the average over a flux surface of the number of transits made around said toroidal space in the toroidal direction by a magnetic flux line in making a single transit in the poloidal direction, and the absolute magnitude of q being less than 1 within said plasma current. means for producing net applied toroidal magnetic field in said toroidal space. a confinement vessel having an axis and an axial direction and walls surrounding the axis thereof for defining a space and confining gas therein, means for producing substantial plasma current within said space in said axial direction, said plasma current producing a substantial magnetic field around said axis within the plasma, first and second windings wound substantially helically around said vessel with the same sense of twist at substantially the same pitch, said first windings and said second windings being disposed alternately and substantially equally spaced around said axis, and means for passing first direct current through said first windings and second direct current through said second windings in the direction counter to said first direct current to generate a helical magnetic field acting in combination with the magnetic field produced by said plasma current to produce nested magnetic flux surfaces within said plasma, which surfaces are spaced from said vessel walls and are closed in the direction transverse of the axis, wherein a safety factor q within said plasma current is the sum of two components, one being axisymmetric and substantially proportional to the ratio of axial magnetic field to circumaxial magnetic field, and the other being nonaxisymmetric and substantially helically symmetric and substantially the quantity ##EQU47## where b is a measure of the strength of the magnetic field from said helical windings, L is the length of said plasma current within said space, r is the average radius of the magnetic flux surface, B.sub..theta.,o is the circumaxial magnetic field produced by said axial plasma current, ##EQU48## is the partial derivative of B.sub..theta.,o with respect to r, l is the number of said first windings, k is the wavenumber of the magnetic field produced by said first and second windings, I.sub.l (kr) is the modified Bessel function of order l, and I.sub.l '(kr) is the derivative of I.sub.l (kr) with respect to its argument, q being defined as the average over a flux surface of the number of transits of the length of the plasma current by a magnetic flux line in making a single transit around the plasma current, and the absolute magnitude of q being less than 1 said plasma current. a confinement vessel for defining a space having an axis and an axial direction and confining gas therein, means for producing substantial plasma current within said space in said axial direction, said plasma current producing a substantial magnetic field around said axis, first and second windings wound substantially helically around said vessel with the same sense of twist at substantially the same pitch, said first windings and said second windings being disposed alternately and substantially equally spaced around said vessel, and means for passing first direct current through said first windings and second direct current through said second windings in the direction counter to said first direct current to generate a helical magnetic field acting in combination with said magnetic field produced by said plasma current to produce a variation in a safety factor q with distance from the center of said plasma current in any direction transverse to said plasma current whereby the polarity of q reverses near the boundary of said plasma current, q being defined as the average over a flux surface of the number of transits of the length of the plasma current by a magnetic flux line in making a single transit around the plasma current, wherein the safety factor q within said plasma current is the sum of two components, one being axisymmetric and substantially proportional to the ratio of axial magnetic field to circumaxial magnetic field, and the other being nonaxisymmetric and substantially helically symmetric and substantially the quantity ##EQU49## where b is a measure of the strength of the magnetic field from said helical windings, L is the length of said plasma current within said space, r is the average minor radius of the magnetic flux surface, B.sub..theta.,o is the circumaxial magnetic field produced by said axial plasma current, ##EQU50## is the partial derivative of B.sub..theta.,o with respect to r, l is the number of said first windings, k is the wavenumber of the magnetic field produced by said first and second windings, I.sub.l (kr) is the modified Bessel function of order l, and I.sub.l '(kr) is the derivative of I.sub.l (kr) with respect to its argument, and the absolute magnitude of q being less than 1 within said plasma current. means for producing net applied magnetic field in said direction in said space. inducing substantial toroidal plasma current in said toroidal space, said toroidal plasma current producing a substantial poloidal magnetic field, and generating a helical magnetic field by passing first and second direct currents through respective first and second windings wound substantially helically around said vessel with the same sense of twist at substantially the same pitch, said first windings and said second windings being disposed alternately and substantially equally spaced around the minor circumference of said vessel, said second direct current being passed in the direction counter to said first direct current, said helical magnetic field combining with said poloidal magnetic field to produce closed and nested magnetic flux surfaces spaced from said vessel walls, wherein a safety factor q within said plasma current is the sum of two components, one being axisymmetric and substantially proportional to the ratio of toroidal magnetic field to poloidal magnetic field, and the other being nonaxisymmetric and substantially helically symmetric and substantially the quantity ##EQU51## where b is a measure of the strength of the magnetic field from the helical windings, R is the major radius of said toroidal space, r is the average minor radius of the magnetic flux surface, B.sub..theta.,o is the poloidal magnetic field produced by said toroidal plasma current, ##EQU52## is the partial derivative of B.sub..theta.,o with respect to r, l is the number of said first windings, k is the wavenumber of the magnetic field produced by said first and second windings, I.sub.l (kr) is the modified Bessel function of order l, and I.sub.l '(kr) is the derivative of I.sub.l (kr) with respect to its argument, q being defined as the average over a flux surface of the number of transits made around said toroidal space in the toroidal direction by a magnetic flux line in making a single transit in the poloidal direction, and the absolute magnitude of q being less than 1 within said plasma current. inducing substantial toroidal plasma current in said toroidal space, said toroidal plasma current producing a substantial poloidal magnetic field, and generating a helical magnetic field by passing first and second direct currents through respective first and second windings wound substantially helically around said vessel with the same sense of twist at substantially the same pitch, said first windings and said second windings being disposed alternately and substantially equally spaced around the minor circumference of said vessel, said second direct current being passed in the direction counter to said first direct current, said helical magnetic field combining with said poloidal magnetic field to produce a variation in a safety factor q with minor radius at any poloidal angle whereby the polarity of q reverses near the outer edge of the plasma current, q being defined as the average over a flux surface of the number of transits made around said toroidal space in the toroidal direction by a magnetic flux line in making a single transit in the poloidal direction, and the absolute magnitude of q being less than 1 within said plasma current. inducing substantial plasma current in said toroidal space, said toroidal plasma current producing a substantial poloidal magnetic field, generating a vertical magnetic field within said toroidal space, said vertical field being perpendicular to the equatorial plane of said toroidal space, and generating a helical magnetic field by passing first and second direct currents through respective first and second windings wound substantially helically around said vessel with the same sense of twist at substantially the same pitch, said first windings and said second windings being disposed alternately and substantially equally spaced around the minor circumference of said vessel, said second direct current being passed in the direction counter to said first direct current, said helical magnetic field combining with said poloidal magnetic field and said vertical field to produce closed and nested magnetic flux surfaces spaced from vessel walls, wherein a safety factor q within said plasma current is the sum of two components, one being axisymmetric and substantially proportional to the ratio of toroidal magnetic field to poloidal magnetic field, and the other being nonaxisymmetric and substantially helically symmetric and substantially the quantity ##EQU53## where b is a measure of the strength of the magnetic field from the helical windings, R is the major radius of said toroidal space, r is the average minor radius of the magnetic flux surface, B.sub..theta.,o is the poloidal magnetic field produced by said toroidal plasma current, ##EQU54## is the partial derivative of B.sub..theta.,o with respect to r, l is the number of said first windings, k is the wavenumber of the magnetic field produced by said first and second windings, I.sub.l (kr) is the modified Bessel function of order l, and I.sub.l '(kr) is the derivative of I.sub.l (kr) with respect to its argument, q being defined as the average over a flux surface of the number of transits made around said toroidal space in the toroidal direction by a magnetic flux line in making a single transit in the poloidal direction, and the absolute magnitude of q being less than 1 within said plasma current. producing substantial plasma current in said space in said axial direction, said plasma current producing a substantial magnetic field around said axis, and generating a helical magnetic field by passing first and second direct currents through respective first and second windings wound substantially helically around said vessel with the same sense of twist at substantially the same pitch, said first windings and said second windings being disposed alternately and substantially equally spaced around said axis, said second direct current being passed in the direction counter to said first direct current, said helical magnetic field combining with said magnetic field produced by said plasma current to produce nested magnetic flux surfaces within said space, which surfaces are spaced from said vessel walls and are closed in the direction transverse of the axis, wherein a safety factor q within said plasma current is the sum of two components, one being axisymmetric and substantially proportional to the ratio of axial magnetic field to circumaxial magnetic field, and the other being nonaxisymmetric and substantially helically symmetric and substantially the quantity ##EQU55## where b is a measure of the strength of the magnetic field from said helical windings, L is the length of said plasma current within said space, r is the average radius of the magnetic flux surface, B.sub..theta.,o is the circumaxial magnetic field produced by said axial plasma current, ##EQU56## is the partial derivative of B.sub..theta.,o with respect to r, l is the number of said first windings, k is the wave number of the magnetic field produced by said first and second windings, I.sub.l (kr) is the modified Bessel function of order l, and I.sub.l (kr) is the derivative of I.sub.l (kr) with respect to its argument, q being defined as the average over a flux surface of the member of transits of the length of the plasma current by a magnetic flux line in making a single transit around the plasma current, and the absolute magnitude of q being less than 1 within said plasma current. producing substantial plasma current in said space in said axial direction, said plasma current producing a substantial magnetic field around said axis, and generating a helical magnetic field by passing first and second direct currents through respective first and second windings wound substantially helically around said vessel with the same sense of twist at substantially the same pitch, said first windings and said second windings being disposed alternately and substantially equally spaced around said vessel, said second direct current being passed in the direction counter to said first direct current, said helical magnetic field combining with said magnetic field produced by said plasma current to produce a variation in a safety factor q with distance from the center of said plasma current in any direction transverse to said plasma current whereby the polarity of q reverses near the boundary of said plasma current, q being defined as the average over a flux surface of the number of transits of the length of the plasma current by a magnetic flux line in making a single transit around the plasma current, and the absolute magnitude of q being less than 1 within said plasma current. 2. A toroidal plasma device according to claim 1 wherein the sense of twist of said first and second windings and the direction of said plasma current produce a variation in the safety factor q with minor radius at any poloidal angle whereby the polarity of q reverses near the outer edge of said plasma current. 3. A toroidal plasma device comprising 4. A toroidal plasma device comprising 5. A toroidal plasma device according to claim 4 wherein said means for generating a vertical field comprises a plurality of circular coils coaxial with the major axis of said toroidal space. 6. A toroidal plasma device according to claim 4 wherein the sense of twist of said first and second windings and the direction of said plasma current produce a variation in the safety factor q with minor radius at any poloidal angle whereby the polarity of q reverses near the outer edge of said plasma current. 7. A toroidal plasma device according to claim 6 wherein said means for generating a vertical field comprises a plurality of circular coils coaxial with the major axis of said toroidal space. 8. A toroidal plasma device according to claim 3 including means for generating a vertical field within said toroidal space, said vertical field being perpendicular the equatorial plane of said toroidal space. 9. A toroidal plasma device according to any one of claims 1, 2, 3, 4, 5, 6, 7 and 8 further comprising 10. A toroidal plasma device according to claim 9 wherein said helical magnetic field and said poloidal magnetic field generate a separatrix within said toroidal space bounding the region in which said nested magnetic flux surfaces exist. 11. A toroidal plasma device according to claim 10 wherein the distance of said separatrix from the minor axis of said toroidal space increases with an increase in said plasma current. 12. A toroidal plasma device according to claim 9 including means for separating said plasma current from said vessel walls. 13. A toroidal plasma device according to any one of claims 1, 2 and 3 wherein said first and second direct currents are substantially equal. 14. A toroidal plasma device according to claim 9 wherein said means for producing net applied toroidal magnetic field comprises means for unbalancing said first and second direct currents. 15. A toroidal plasma device according to claim 9 wherein said means for generating magnetic flux comprises a solenoid coaxial with the major axis of said toroidal space. 16. A toroidal plasma device according to claim 15 wherein said toroidal confinement vessel comprises a conductive thin toroidal wall. 17. A toroidal plasma device according to claim 16 including means for evacuating said toroidal space. 18. A toroidal plasma device according to claim 17 wherein the number of said first and second windings around the minor circumference are each two. 19. A toroidal plasma device according to claim 17 wherein the number of said first and second windings around the minor circumference are each three. 20. A toroidal plasma device according to claim 17 including a toroidal shell containing and spaced from said confinement vessel, said shell being of electrically conductive material with the conductive path interrupted in the toroidal direction. 21. A toroidal plasma device according to claim 9 wherein said first and second windings are wound at such pitch as to produce relatively small interwinding forces when said first and second direct currents are passed therethrough. 22. A toroidal plasma device according to claim 21 wherein said windings are wound at an angle of approximately 45.degree. to the minor axis of the confinement vessel. 23. A plasma device comprising 24. A plasma device according to claim 23 wherein the sense of twist of said first and second windings and the direction of said plasma current produce a variation in the safety factor q with distance from the center of said plasma current in any direction transverse to said plasma current whereby the polarity of q reverses near the boundary of said plasma current. 25. A plasma device according to either one of claims 23 and 24 wherein said helical magnetic field and said magnetic field produced by said plasma current generate a separatrix within said space bounding the region in which said nested magnetic flux surfaces exist. 26. A plasma device according to claim 25 wherein the distance of said separatrix from the center of said plasma current increases with an increase in said plasma current. 27. A plasma device according to either one of claims 23 and 24 including means for separating said plasma current from said vessel walls. 28. A plasma device comprising 29. A plasma device according to any one of claims 23, 24 and 28 wherein said first and second direct currents are substantially equal. 30. A plasma device according to any one of claims 23, 24 and 28 further comprising 31. A plasma device according to claim 30 wherein said means for producing net applied magnetic field comprises means for unbalancing said first and second direct currents. 32. A plasma device according to claim 30 wherein said first and second windings are wound at such pitch as to produce relatively small interwinding forces when said first and second direct currents are passed therethrough. 33. A plasma device according to claim 32 wherein said windings are wound at an angle of approximately 45.degree. to the axial direction. 34. A method of operating a toroidal plasma device having a toroidal confinement vessel with walls defining a toroidal space and confining gas therein, said method comprising 35. A method according to claim 34 wherein the sense of twist of said first and second windings and the direction of said plasma current produce a variation in the safety factor q with minor radius at any poloidal angle whereby the polarity of q reverses near the outer edge of said plasma current. 36. A method of operating a toroidal plasma device having a toroidal confinement vessel with walls defining a toroidal space and confining gas therein, said method comprising 37. A method of operating a toroidal plasma device having a toroidal confinement vessel with walls defining a toroidal space and confining gas therein, said method comprising 38. A method according to claim 37 wherein said vertical field is generated by currents in circular windings coaxial with the major axis of said toroidal space. 39. A method according to claim 37 wherein the sense of twist of said first and second windings and the direction of said plasma current produce a variation in the safety factor q with minor radius at any poloidal angle whereby the polarity of q reverses near the outer edge of said plasma current. 40. A method according to claim 39 wherein said vertical field is generated by currents in circular windings coaxial with the major axis of said toroidal space. 41. A method according to claim 36 including the step of generating a vertical magnetic field within said toroidal space, said vertical field being perpendicular to the equatorial plane of said toroidal space. 42. A method according to any one of claims 34, 35, 36, 37, 38, 39, 40 and 41 wherein net toroidal magnetic field is applied in said toroidal space. 43. A method according to claim 42 wherein said helical magnetic field and said poloidal magnetic field generate a separatrix within said toroidal space bounding the region in which said nested magnetic flux surfaces exist. 44. A method according to claim 43 wherein the distance of said separatrix from the minor axis of said toroidal space is increased by increasing said plasma current. 45. A method according to claim 42 wherein said plasma current is separated from said vessel walls. 46. A method according to any one of claims 34, 35 and 36 wherein said fist and second direct currents are substantially equal. 47. A method according to claim 42 wherein said net applied toroidal magnetic field is produced at least in part by the unbalance of said first and second direct currents. 48. A method according to claim 42 wherein said plasma current is induced by changing magnetic flux linking said toroidal space. 49. A method of operating a plasma device having a confinement vessel with an axis and an axial direction with walls surrounding the axis thereof for defining a space and confining gas therein, said method comprising 50. A method according to claim 49 wherein the sense of twist of said first and second windings and the direction of said plasma current produce a variation in the safety factor q with distance from the center of said plasma current in any direction transverse to said plasma current whereby the polarity of q reverses near the boundary of said plasma current. 51. A method according to either one of claims 49 and 50 wherein said helical magnetic field and said magnetic field produced by said plasma current generate a separatrix within said space bounding the region in which said nested magnetic flux surfaces exist. 52. A method according to claim 51 wherein the distance of said separatrix from the center of said plasma current is increased by increasing said plasma current. 53. A method according to either one of claims 49 and 50 wherein said plasma current is separated from said vessel walls. 54. A method of operating a plasma device having a confinement vessel for defining a space having an axis and an axial direction and confining gas therein, said method comprising 55. A method according to any one of claims 49, 50 and 54 wherein said first and second direct currents are substantially equal. 56. A method according to any one of claims 49, 50 and 54 wherein net magnetic field is applied in the direction of plasma current. 57. A method according to claim 56 wherein said net applied magnetic field is produced at least in part by the unbalance of said first and second direct currents. 58. A method according to claim 56 wherein said plasma current is induced by changing magnetic flux linking said space. 59. A toroidal plasma device according to any one of claims 1, 2 and 4 to 7 wherein said helical magnetic field and said poloidal magnetic field generate a separatrix within said toroidal space bounding the region in which said nested magnetic flux surfaces exist. 60. A toroidal plasma device according to claim 59 wherein the distance of said separatrix from the minor axis of said toroidal space increases with an increase in said plasma current. 61. A toroidal plasma device according to any one of claims 1, 2 and 4 to 7 including means for separating said plasma current from said vessel walls. 62. A method according to any one of claims 34, 35 and 37 to 40 wherein said helical magnetic field and said poloidal magnetic field generate a separatrix within said toroidal space bounding the region in which said nested magnetic flux surfaces exist. 63. A method according to claim 62 wherein the distance of said separatrix from the minor axis of said toroidal space is increased by increasing said plasma current. 64. A method according to any one of claims 34, 35 and 37 to 40 wherein said plasma current is separated from said vessel walls. 65. A plasma device according to any one of claims 1, 2, 3, 23, 24, 28, 4, 5, 6 and 7 wherein the number of said first and second windings are each more than 1. 66. A toroidal plasma device according to claim 9 wherein the number of said first and second windings are each more than 1. 67. A plasma device according to claim 30 wherein the number of said first and second windings are each more than 1. 68. A method according to any one of claims 34, 35, 36, 49, 50, 54, 37, 38, 39 and 40 wherein the number of said first and second windings are each more than 1. 69. A method according to claim 42 wherein the number of said first and second windings are each more than 1. 70. A method according to claim 56 wherein the number of said first and second windings are each more than 1.
	description	1. Field of the Invention The present invention relates to a multi-leaf collimator that narrows the radiation field emitted to an object and a radiotherapy unit provided with the same. 2. Description of the Related Art Radiation therapy has been widely used as treatment for affected parts such as cancer and tumors. This radiation therapy is treatment with the destruction of tissue cells in an affected part, inhibition of cell division, or the like by emitting radiation to the affected part. It is important to precisely irradiate an affected part in such radiation therapy. By precisely irradiating an affected part, damage to normal cells around an affected part can be minimized, thereby enabling effective irradiation to an affected part. In order to ensure precise irradiation, generated radiation is narrowed with a diaphragm part or multi-leaf collimator so that the shape of the radiation field becomes similar to that of the affected part. A multi-leaf collimator comprises a plurality of leaf blocks and a movable mechanism paired with these leaf blocks. The leaf blocks are made of a material such as tungsten that absorbs radiation. By moving the leaf blocks between a radiation source and sites where there is no need to emit radiation, a radiation field is narrowed to a predetermined shape (e.g., refer to Japanese Published Unexamined Application No. 1999-216197). This movable mechanism is shown in FIG. 1. As shown in FIG. 1, the movable mechanism is provided with a drive motor M, which is a motor with a reducer, and an output gear OG. The leaf blocks are moved with the drive power transmitted from the output gear OG. A plurality of transmission gears TG are arranged between the drive motor M and output gear OG, respectively rotating as they engage with each other, thereby transmitting drive power from the drive motor M to the output gear OG. In this movable mechanism, location detection part such as an encoder E or potentiometer P are placed on one of the multiple transmission gears TG. Displacement of the leaf blocks is correlated with the amount of rotation of the drive motor M, output gear OG, and the plurality of transmission gears TG. The encoder E and potentiometer P detect the locations of the leaf blocks by detecting the rotation of the transmission gears TG. Furthermore, although the encoder E and potentiometer P have the same function, they both may be arranged for redundancy. Thus, if a movable mechanism comprises a plurality of gears as above, backlash occurs at each of the gears. With the method of detecting the locations of the leaf blocks based on rotation of one of the multiple gears, errors due to backlash are observed between the detected locations of the leaf blocks and the actual locations thereof. It is possible to minimize errors caused by backlash, such as by correcting the detected locations of the leaf blocks with the backlash of each gear. However, gears engage with each other physically, so the tops of the teeth become worn over time. As this wear progresses over time, errors caused by backlash will increase. Therefore, even if the locations are corrected with pre-calculated backlash, errors will gradually increase, resulting in decreased accuracy of specifying the locations of the leaf blocks over time. It is preferable to keep errors in shape between an affected part and a radiation field to within 1 mm around the location of the affected part. Thus, the difference in locations of the leaf blocks must be kept to approximately 0.3 mm. This is based on the ratio of distance between the surface of an installed multi-leaf collimator from a radiation source and the location of an affected part therefrom. Therefore, errors in detected locations caused by backlash affect therapeutic effectiveness to a considerable degree. This invention is intended to provide a multi-leaf collimator that avoids the effect of backlash in detecting the displacement or locations of leaf blocks and that precisely detects the displacement or locations of leaf blocks and to provide the radiotherapy unit provided with the same. In the first aspect of this invention, a multi-leaf collimator that narrows a radiation field to a predetermined shape is provided with leaf blocks movable in the direction of the radiation field and a detecting element that detects the displacement or locations of the leaf blocks. The leaf blocks are provided with pattern images drawn along the direction of movement on a predetermined surface. Moreover, the detecting element acquires an image of fixed-point via fixed-point observation in the direction of said predetermined surface and detects the displacement or locations of said leaf blocks based on the arranged locations of the pattern images existing in this image of fixed-point. In the second aspect of this invention, a radiotherapy unit is provided with a radiation source for irradiating radiation, a bed on which to place the object, and a multi-leaf collimator between the radiation source and the bed that narrows a radiation field irradiated from said radiation source to a predetermined shape. The multi-leaf collimator is provided with leaf blocks movable in the direction of the radiation field and a detecting element that detects the displacement or locations of the leaf blocks. The leaf blocks are provided with pattern images drawn along the direction of movement on a predetermined surface. Furthermore, the detecting element acquires an image of fixed-point via fixed-point observation in the direction of said predetermined surface and detects the displacement or locations of the leaf blocks based on the arranged locations of the pattern images existing in this image of fixed-point. According to the first and second aspects 1 and 2 of this invention, the displacement and locations of the leaf blocks can be detected without making contact, and errors due to the effect of backlash and gear wear in detecting the displacement and the locations can be prevented. Therefore, regardless of backlash, the locations of leaf blocks can be detected with high precision, and the radiation field can be matched to the shape of the affected part with high precision. FIG. 2 is an external view showing a radiotherapy unit 1 related to this embodiment. The radiotherapy unit 1 is an apparatus for treating the affected part of an object P. The radiotherapy unit 1 treats the affected part by emitting radiation at the affected part of the object P. This radiotherapy unit 1 is provided with a radiation head 2 that generates radiation and a bed 5 on which the object P is placed. The radiation head 2 and bed 5 are arranged facing each other. Radiation generated in the radiation head 2 is emitted in the direction toward the bed 5. This radiotherapy unit 1 is fixed with a fixing gantry 3 on the surface on which the apparatus is installed. A rotating gantry 4 is supported by this fixing gantry 3 in the air. The radiation head 2 is installed on this rotating gantry 4. The rotating gantry 4 is a rough L-shaped steric figure. The rotating gantry 4 has an arm 4a supported by the fixing gantry 3, and the radiation head 2 is installed on the other arm 4b. The radiation head 2 is facing the direction of the bed S. In radiation therapy, it is necessary to precisely match the affected part of the object P placed on the bed 5 and the isocenter of radiation. Therefore, the rotating gantry 4 is supported by the fixing gantry 3 through a turning shaft 3a. By turning the rotating gantry 4 around the turning shaft 3a, the direction of the radiation head 2 will change. By changing the direction of the radiation head 2, the radiation to be generated is emitted at different angles, so the isocenter can be changed around the turning shaft 3a. Furthermore, the bed 5 is movable in the direction of the body axis of the object P, the direction of the radial axis of radiation, and the direction of rotation to remain parallel with the installation surface of the radiotherapy unit 1. Movement in these directions allows the location of the isocenter to change, matching the isocenter with the affected part in combination with the rotation of the rotating gantry 4. FIG. 3 is a view showing the basic structure of the radiotherapy unit 1. As shown in FIG. 3, a radiation source 21, a diaphragm part 22 and a multi-leaf collimator 23 are placed within the radiation head (head) 2 collaterally in the direction toward the bed 5. The radiation source 21 comprises an electron accelerator, the target of the electron beam, and so on. Radiation is generated by accelerating electrons with the electron accelerator and colliding them against the target of the electron beam (target). Radiation that is generated may be a photon beam (X-ray and γ-ray, etc.), electron beam, heavy particle beam (proton, helium, carbon, neon, π meson beam, neutron ray, and so on), and so forth. The diaphragm part 22 and the multi-leaf collimator 23 are arranged in the range of radiation, narrowing the range of radiation and creating a radiation field F that matches the shape of the affected part. The diaphragm part 22 comprises a block pair facing each other over the radiation axis. The material of the block pair possesses a property of absorbing radiation such as tungsten. This block pair narrows the range of radiation by bringing them together or separating them. FIG. 4 is a perspective view showing the multi-leaf collimator 23. The multi-leaf collimator 23 comprises leaf blocks 23A and 23B facing each other over the radiation axis. The leaf blocks 23A and 23B have tapered cross-sectional surfaces and arch-like side surfaces with short perimeters. Moreover, the material possesses a property of absorbing radiation such as tungsten. A plurality of pairs of leaf blocks 23A and 23B are placed close to or in contact with each other in the direction of the side surfaces. A movable mechanism 24 is placed on each of the leaf blocks 23A and each of the leaf blocks 23B. The movable mechanism 24 displaces each of the leaf blocks 23A or 23B targeted for movement with the same circular orbit, with the radiation source 21 as the center of the direction of movement. With this movable mechanism 24, pairs of leaf blocks 23A and 23B are moved toward or separate from each other, narrowing the radiation field F to the appropriate shape. Radiation emitted to places other than the radiation field F is absorbed by the leaf blocks 23A and 23B, and only radiation that passes through the radiation field F will pass through the multi-leaf collimator 23. Furthermore, a detecting element 25 is placed on each of the leaf blocks 23A and each of the leaf blocks 23B. The detecting element 25 detects displacement of the leaf blocks 23A or 23B respectively targeted for detection. Displacement indicates the vector for showing changes in locations, including direction of movement and distance. The result of displacement detection with the detecting element 25 is fed back to the movable mechanism 24. The multi-leaf collimator 23 displaces each of the leaf blocks 23A and 23B to the intended location according to the feedback of the result of displacement detection, and creates a radiation field F that matches the shape of the affected part. Differences between the accumulated results of displacement detection and intended locations are calculated, and each of the leaf blocks 23A and 23B is displaced so as to eliminate the differences. A simplified operation of the radiotherapy unit 1 will be described below based on FIG. 5. FIG. 5 is a flowchart showing the simplified operation of the radiotherapy unit 1. First, with the radiotherapy unit 1, according to the predefined treatment plan, the rotating gantry 4 is rotated, and the bed 5 is moved in the direction of the body axis of the object P, moved in the direction of the radial axis of radiation, and moved in the direction of rotation to remain parallel with the installation surface of the radiotherapy unit 1 in order to match the isocenter of radiation with the location of the affected part by these movements (S01). Secondly, the movable mechanism 24 is driven, and each pair of leaf blocks 23A and 23B is approached to or separated from each other (S02). The detecting element 25 detects displacement of each of the leaf blocks 23A or 23B targeted for detection (S03). By accumulating information regarding detected displacement, the location of each of the leaf blocks 23A and 23B is specified (S04). The setup location of each of the leaf blocks 23A and 23B included in the treatment plan and the specified current location of each of the leaf blocks 23A and 23B are compared (S05), and if the current location does not reach the setup location (S06, No), return to S02, and each of the leaf blocks 23A and 23B will be further moved toward or separated from each other. If the current location reaches the setup location (S06, Yes), displacement of each of the leaf blocks 23A and 23B with the movable mechanism 24 is stopped (S07). Subsequently, radiation is generated in the radiation source 21 with radiation dosage and emission time conforming to the treatment plan (S08), and emitted at the affected part of the object P. The multi-leaf collimator 23 of such a radiotherapy unit 1 will be described in further detail. FIG. 6 is a side view of the multi-leaf collimator 23, and FIG. 7 is a cross-sectional view of the leaf block 23B cut in the direction of the side surface, particularly showing the periphery around the bottom surface. Only the structure of the leaf blocks 23B is described below, as the structure of the leaf blocks 23A is the same. As shown in FIG. 6 and FIG. 7, a toothed wheel cutting 232 is drawn along the longitudinal direction on the outer circumferential arc surface of the leaf block 23B. The movable mechanism 24 is provided with a drive motor and an output gear 24A. The toothed wheel cutting (gear cutting) 232 and output gear 24A engage with each other. Between the drive motor and output gear 24A, the drive power is transmitted from the drive motor to the output gear 24A through a plurality of gears. The movable mechanism 24 rotates the multiple gears by activating the drive motor, and transmits drive power to the output gear 24A. This drive power rotates the output gear 24A. The leaf block 23B moves in the direction of movement to narrow or broaden the radiation field F along with rotation of the output gear 24A through engagement of the output gear 24A and the toothed wheel cutting 232. Moreover, pattern images 231 are drawn along the longitudinal direction, or in other words along the direction of movement of the leaf block 23B, on the outer circumferential arc surface of the leaf block 23B. A plurality of the pattern images 231 are consecutively drawn. These pattern images 231 extend through the same range as the movable range of the leaf block 23B. In other words, the toothed wheel cutting 232 and pattern images 231 are drawn in parallel on the outer circumferential arc surface of the leaf block 23B. The detecting element 25 is provided with an irradiation part 26 such as a light-emitting diode (diode), an image sensor 27 such as CCD, and displacement acquisition part 28 comprising a plurality of processing circuits or the like as shown in FIG. 6 and FIG. 7. The irradiation part 26 irradiates a beam at a fixed point. The direction of irradiation from the irradiation part 26 is fixed, so the fixed point is an absolute location, independent from displacement of the leaf block 23B, and has a predetermined area. The irradiation part 26 sets the fixed point in the direction of the outer circumferential arc surface of the leaf block 23B so that the fixed point includes part of the region where the pattern images 231 are drawn. The image sensor 27 performs fixed-point observations in the direction of the outer circumferential arc surface of the leaf block 23B at certain intervals. The observation point is the fixed point where the irradiation part 26 irradiates a beam. The image sensor 27 is subject to the beam reflected from this fixed point and acquires an image of fixed-point 31 at specified time intervals. The image of fixed-point 31 is an image in the fixed-point region. One of the pattern images 231 drawn on the leaf block 23B exists in the acquired image of fixed-point 31. Furthermore, the detecting element 25 is provided with a lens 29 in an optical system between the fixed point and image sensor 27, aligning a scattering beam and adjusting magnification of images acquired by the image sensor 27. The displacement acquisition part 28 analyzes the image of fixed-point 31 and acquires displacement of the leaf block 23B. If the leaf block 23B is displaced, the pattern images 231 drawn on the leaf block 23B will also move. The arranged locations of the pattern images 231 on the image of fixed-point 31 change according to displacement of the leaf block 23B. It recognizes the arranged locations of the pattern images 231, determines the difference between the arranged locations and acquires displacement of the arranged leaf block 23B from the differences of the arranged locations. FIG. 8 is a view showing an example of the pattern images 231 drawn on the leaf block 23B. As shown in FIG. 8, the pattern images 231 have predetermined marks, and a plurality of these are placed in consecutive lines. The patterns of pattern images 231 include specific patterns 231a that are in predetermined locations within the patterns. The specific patterns 231a are characteristic patterns that are different from patterns of other regions (including planes) of the pattern images 231. The image sensor 27 captures fixed-point regions over time where the pattern images 231 are drawn. The specific patterns 231a of the pattern images 231 are included within the image of fixed-point 31. The displacement acquisition part 28 specifies the locations of specific patterns 231a placed within the image of fixed-point 31, and acquires displacement from the differences acquired over time. Therefore, regardless of the amount of displacement of the leaf block 23B at a certain time, it is preferable for some or all of the specific patterns 231a to exist within the image of fixed-point 31. Thus, it is preferable for the size of the pattern images 231 to be smaller than that of the image of fixed-point 31. The pattern images 231 have a specific pattern 231a carved deeper than the other segmented regions in predetermined segmented regions where, for example, an approximately 0.5-mm square region is divided into sixteen segments. For example, grating in the corner is the specific pattern 231a. The size of approximately 0.5 mm square is to be a region smaller than the fixed-point region observed by the image sensor 27. The patterns can be evenly carved, or can be carved by loose grooves. FIG. 9 is a block diagram showing the detecting element 25 in detail, particularly for describing the displacement acquisition part 28 in more detail. The displacement acquisition part 28 is provided with an image recognition part 281, comparison part 282, and converter 284. The image recognition part 281 is electrically connected to the image sensor 27. The image of fixed-point 31 acquired by the image sensor 27 is input into the image recognition part 281. This image recognition part 281 scans the image of fixed-point 31 and recognizes the arranged locations of the pattern images 231 existing in the image. In scanning the image of fixed-point 31, the specific patterns 231a that the pattern images 231 existing within the image of fixed-point 31 form are searched, and arrangement location information showing the arranged locations of the specific patterns 231a is acquired. The arrangement location information is shown in the coordinate range where the specific patterns 231a are located. Furthermore, the leaf block 23B only moves one-dimensionally in a direction to narrow or broaden the radiation field F, so locations of the specific patterns 231a change one-dimensionally. Therefore, the image recognition part 281 can scan only the range of lines that may comprise the specific patterns 231a. Moreover, the arrangement location information can be information that shows the range of rows of coordinates where the specific patterns 231a are distributed. The scanning process detects brightness shown by pixels in the scan range and extracts the coordinates of pixels with a specific luminance value shown by the specific patterns 231a. If the specific patterns 231a is deeply carved or consists of loose grooves, the brightness will be lower than that of the other regions. The comparison part 282 acquires differences in pattern images 231 by comparing the arranged locations of the pattern images 231 existing in two images of fixed-point 31 that differ over time. The comparison part 282 and image recognition part 281 are electrically connected, and the arrangement location information is input into the comparison part 282 from the image recognition part 281. This comparison part 282 is provided with an arrangement location storage 283 comprising a memory circuit. The arrangement location information for pattern images 231 existing in the previously acquired image of fixed-point 31 is stored in the arrangement location storage 283. The comparison part 282 reads the previous arrangement location information stored in the arrangement location storage 283, differentiates it from the arrangement location information newly output by the image recognition part 281, and acquires information regarding the difference. Information regarding the difference shows differences in specific patterns 231a, including information for the direction of difference and amount of difference. The differentiation process differentiates the arrangement location information newly input by the image recognition part 281 from the previous arrangement location information, and acquires the remaining coordinate range. Distribution of the remaining coordinate range shows the amount of difference, and the difference in values between the coordinate range shown by the arrangement location information acquired previously and the remaining coordinate range shows the direction of difference. A value of the remaining coordinate range that is smaller shows that the difference has moved in the direction of the pixels that takes a lower value for the row coordinate. The value of the coordinate range may, for example, be medial coordinates of each point of the coordinate range. The comparison part 282 creates information regarding the difference, including this amount of difference and direction of difference, and outputs it to the converter 284. The converter 284 is electrically connected to the comparison part 282, with information regarding the difference input therein. This converter 284 has a conversion formula for converting information regarding the difference that shows displacement of pixel units into an actual displacement amount of the leaf block 23B. The input information regarding the difference is converted into actual displacement of the leaf block 23B with the conversion formula. Furthermore, displacement of the leaf block 23B detected with the detecting element 25 is input into a location counter 30 provided with the radiotherapy unit 1. The location counter 30 comprises a data storage such as semiconductor memory, accumulatively adding displacement detected with the detecting element 25 and recording it therein. This location counter 30 can comprise a detecting element 25. In the radiotherapy unit 1 of this embodiment, the leaf block 23B is moved to the default location by the movable mechanism 24 at the time of starting the apparatus when displacement starts to be recorded accumulatively, so the cumulative value of displacement shows the location of the leaf block 23B. Based on FIG. 10 and FIG. 11, the detection operation for displacement of the leaf block 23B of this radiotherapy unit 1 will be described. FIG. 10 is a flow chart showing the detection operation for displacement of the leaf block 23B. FIG. 11 is a view showing the consecutive images of fixed-point 31 over time captured by the image sensor 27 at a certain period of time, with (a) being the image of fixed-point 31a before displacement of the leaf block 23B at that period of time and (b) being the image of fixed-point 31b after displacement of the leaf block 23B. First, when displacement of the leaf block 23B starts with the movable mechanism 24, a beam is irradiated at the fixed point from the irradiation part 26 (S21), and the image sensor 27 acquires the images of fixed-point 31 existing at the fixed point at certain intervals (S22). Herein, the image sensor is supposed to acquire the image of fixed-point 31 before displacement of the leaf block 23B at a certain period of time. The image recognition part 281 recognizes the arranged locations of the pattern images 231 existing in the image of fixed-point 31a before this displacement, acquiring arrangement location information D1 of the specific pattern 231a. This arrangement location information D1 is stored in the arrangement location storage 283. Secondly, after displacement of the leaf block 23B at that certain period of time, when the image sensor 27 acquires the image of fixed-point 31b existing at the fixed point, the image recognition part 281 scans the range of lines where the specific pattern 231a existing in the image of fixed-point 31b is placed (S23). While scanning, the image of fixed-point 31b finds pixels with a luminance value of the specific pattern 231a, the coordinates of the pixels are acquired as arrangement location information D2 of the specific pattern 231a (S24). The comparison part 282 reads the arrangement location information D1 from the arrangement location storage 283 (S25). When the arrangement location information D1 is read, the arrangement location information D1 is differentiated from the arrangement location information D2 to acquire the difference (S26). In the displacement shown in FIG. 11, this difference creates information regarding the difference indicating that the specific pattern has moved to the left by a coordinate range D3. When information regarding the difference is created, the arrangement location information D2 of the specific pattern 231 acquired after the displacement is updated and stored in the arrangement location storage 283 as arrangement location information D1 before displacement in the next period of time (S27). Moreover, when information regarding the difference is acquired, the converter 284 converts this information regarding the difference into an actual displacement value for the leaf block 23B using the conversion formula (S28). The displacement value is output to the location counter 30 and is added to the cumulative value showing the location of the leaf block 23B stored by the location counter 30 (S29). If the cumulative value showing the location of the leaf block 23B does not match the value showing the intended setup location of the leaf block 23B (S30, No), S23 to S31 are repeated to detect the difference by comparing the arranged locations of the pattern images 231 between the new image of fixed-point 31b after displacement and the previous image of fixed-point 31a. If the cumulative value showing the location of the leaf block 23B matches the value showing the intended location of the leaf block 23B (S30, Yes), displacement of the leaf block 23B with the movable mechanism 24 is stopped (S31). As described above, in the radiotherapy unit 1 of this embodiment, displacement of leaf block 23B can be detected without making contact, preventing displacement detection errors due to the effect of backlash and gear wear. Therefore, locations of the leaf blocks 23B can be detected with high precision, and the radiation field F can be matched with the shape of the affected part with high precision. Next, Embodiment 2 of the radiotherapy unit 1 of the present invention will be described. Furthermore, the same codes are used for the same structures and same functions as in Embodiment 1, so detailed descriptions are omitted. The radiotherapy unit 1 of this embodiment detects the location of the leaf block 23B with the detecting element 25. FIG. 12 is a view showing the pattern image 231 drawn on the outer circumferential arc surface of the leaf block 23B. In the radiotherapy unit 1 of this embodiment, one pattern image 231 is drawn along the longitudinal direction, or in other words along the direction of movement of the leaf block 23B, on the outer circumferential arc surface of the leaf block 23B. The pattern image 231 is extended across the same width as the movable range of the leaf block 23B. This pattern image 231 has a plurality of location-specific patterns 231b in parallel at predetermined intervals. These location-specific patterns 231b express location-specific patterns. “Location-specific” indicates that a pattern has different characteristics depending on the location where it is placed. The location-specific patterns 231b of this embodiment are striped and have a predetermined width, with the width of the strip varying, depending on the sequence location. FIG. 13 is a block diagram showing the detecting element 25 that detects the location of the leaf block 23B. As shown in FIG. 13, the detecting element 25 is provided with the irradiation part 26, the image sensor 27, an image recognition part 285, and a locating part 286. The image recognition part 285 is electrically connected to the image sensor 27, and the image of fixed-point 31 acquired by the image sensor 27 is input into it. This image recognition part 281 scans the image of fixed-point 31 and recognizes the location of the arranged pattern images 231 existing in the image. While scanning the image of fixed-point 31, the location-specific patterns 231b of the pattern images 231 existing within the image of fixed-point 31 are searched, and arrangement location information of the location-specific patterns 231b is acquired. The arrangement location information is shown in a coordinate range where the location-specific patterns 231b exist. The acquired arrangement location information includes information for the width of the location-specific patterns 231b existing within the image of fixed-point 31 according to the distribution width of the coordinate shown by this arrangement location information, in addition to the information for locations of the arranged location-specific patterns 231b. That is to say, the arrangement location information includes information to specify the location-specific patterns and information for the arrangement locations within the image of fixed-point 31. This scanning is performed only on the range of lines that possibly comprises the location-specific patterns 231b. The scanning process extracts a coordinates of pixels with a specific luminance value shown by the location-specific patterns 231b. The arrangement location information can be information that shows the range of rows of the coordinate where the location-specific patterns 231b are distributed. The locating part 286 specifies the location of the leaf block 23B from the arrangement location information. This locating part 286 is provided with a conversion table (table) 287 comprising a memory circuit. In the conversion table 287, information for the width of the location-specific patterns 231b and information for the location of the leaf block 23B are stored in pairs. The information for the location of the leaf block 23B is information that shows the location of the leaf block 23B when the location-specific patterns 231b corresponding to the paired information for width is located in the center of the image of fixed-point 31. Locating the leaf block 23B consists of a specific process for the pattern of the location-specific patterns 231b existing within the image of fixed-point 31, a tentative specific process for the location of the leaf block 23B, a process for calculating the difference between the center of the image of fixed-point 31 and the location-specific patterns 231b, and a final locating process for the tentatively specified location of the leaf block 23B. In the specific process for the pattern and the tentative specific process for the location, once the arrangement location information is input, the locating part 286 acquires the distribution width of the coordinate from the arrangement location information, searches information from the conversion table 287 for the width that matches the distribution width, and acquires the location information. In the process of calculating the difference, the locating part 286 calculates the difference between the center coordinates of the image of fixed-point 31 preliminarily stored by the locating part 286 and the medial coordinates of the coordinate range shown by the arrangement location information. In the final locating process, the locating part 286 adds the difference calculated in the process of calculating the difference to the acquired information for the locations. The result of adding the difference to this information for the location is led to the precise location of the leaf block 23B. The detection operation for the location of the leaf block 23B regarding the radiotherapy unit 1 related to this embodiment will be described based on FIG. 14 and FIG. 15. First, when displacement of the leaf block 23B starts with the movable mechanism 24, a beam is irradiated at the fixed point from the irradiation part 26 (S41), and the image sensor 27 acquires the images of fixed-point 31 existing at the fixed point at certain intervals (S42). Secondly, after displacement of the leaf block 23B at a certain period of time, when the image sensor 27 acquires the image of fixed-point 31b existing at the fixed point, the image recognition part 281 scans the range of lines where the location-specific patterns 231b existing in the image of fixed-point 31 are placed (S43). When scanning the image of fixed-point 31 finds pixels with a luminance value of the location-specific patterns 231b, the coordinates of the pixels are acquired as arrangement location information D4 of the location-specific patterns 231b (S44). The locating part 286 extracts the distribution width of the coordinate shown by the arrangement location information D4 (S45), searches the information for the width that shows the same width as this distribution width from the conversion table 287, and acquires the information for the location of the paired leaf blocks 23B (S46). Moreover, the locating part 286 calculates the median of the coordinate range shown by the arrangement location information D4 (S47), and calculates the value of the difference D5 By differing the arrangement location information D4 from the center coordinates of the image of fixed-point 31 that the locating part 286 previously had (S48). The location when the image of fixed-point 31 of the leaf block 23B was acquired is detected by adding the value of the difference D5 to the information for the location acquired from the conversion table 287 (S49). When the acquired location of the leaf block 23B matches the intended location of the leaf block 23B (S50, Yes), displacement of the leaf block 23B with the movable mechanism 24 is stopped (S51). As described above, in the radiotherapy unit 1 of this embodiment, the location of leaf block 23B can be detected without making contact, and location detection errors due to the effect of backlash and gear wear can be prevented. Therefore, locations of the leaf blocks 23B can be detected with high precision, and the radiation field F can be matched to the shape of the affected part with high precision. Next, Embodiment 3 of the radiotherapy unit 1 of the present invention will be described. The same codes are used for the same structures and same functions as Embodiments 1 and 2, so detailed descriptions are omitted. The radiotherapy unit 1 of this embodiment detects displacement and detects locations in a similar manner as in Embodiments 1 or 2. FIG. 16 is a view showing locations of the arranged irradiation part 26 and the image sensor 27. In the radiotherapy unit 1 of this embodiment, a pair of the irradiation part 26 and the image sensor 27 for detecting the displacement or location of the same leaf block 23B and a pair of the irradiation part 26 and the image sensor 27 corresponding to the adjacent leaf block 23B, are differently arranged in lines. As shown in FIG. 16(a), for example, pairs of the irradiation parts 26 and the image sensors 27 corresponding to each of the leaf blocks 23B are arranged alternately in two lines so that the adjacent pairs of the irradiation parts 26 and the image sensors 27 do not belong to the same lines. Furthermore, as shown in FIG. 16(b), for example, pairs of the irradiation parts 26 and the image sensors 27 corresponding to each of the leaf blocks 23B are arranged in three lines so that pairs of the irradiation parts 26 and the image sensors 27 respectively corresponding to the three leaf blocks 23B aligned in sequence all belong to different lines. In order to more precisely match the radiation field F created by the multi-leaf collimator 23 with the shape of the affected part, it is preferable for the thickness of the leaf blocks 23A and 23B to be thinner and for more pairs of leaf blocks 23A and 23B to be placed. In the radiotherapy unit 1 of this embodiment, the adjacent pairs of the irradiation parts 26 and the image sensors 27 are differently arranged in lines, so when a certain irradiation part 26 irradiates radiation, the adjacent image sensors 27 are not subject to any of the borrowed light reflected on the leaf blocks 23B. This enables the acquisition of a clear image of fixed-point 31 without noise in the image of fixed-point 31 acquired by the image sensors 27, enhancing the accuracy of the arrangement location information acquired by the image recognition part 281 or 285, and detecting the displacement or locations of the leaf blocks 23B with higher accuracy. Next, Embodiment 4 of the radiotherapy unit 1 of the present invention will be described. The same codes are used for the same structures and same functions as in Embodiments 1 and 2, so detailed descriptions are omitted. The radiotherapy unit 1 of this embodiment detects displacement and locations in a similar way to Embodiments 1 or 2. The radiotherapy unit 1 of this embodiment is a modified example of improvement in accuracy of the arrangement location information related to Embodiment 3. FIG. 17 is a side view of the leaf block 23B in this embodiment. The irradiation part 26 placed corresponding to each of the leaf blocks 23B irradiates a beam with a wavelength different from the wavelength of the beam irradiated by the adjacent irradiation part 26. The detecting element 25 is provided with an optical filter (filter) 271 in the optical filter between the irradiation part 26 and the image sensor 27. This optical filter 271 only transilluminates a beam with a specific wavelength. The beam with a specific wavelength that is transilluminatable is a beam with a wavelength irradiated by the irradiation part 26 that forms the optical system where the optical filter 271 is placed. This beam irradiated by the irradiation part 26 is transilluminated, and other beams irradiated from the other irradiation parts 26 are absorbed and not transilluminated. In such detecting element 25, the optical filter 271 placed in the optical system formed by the predetermined irradiation part 26 transilluminates a beam with a specific wavelength irradiated by this predetermined irradiation part 26, and has the image sensor 27 subject to the beam. At the same time, when a beam irradiated by the irradiation part 26 placed corresponding to the adjacent leaf blocks 23B enters by scattering on the surface of the adjacent leaf blocks 23B, the optical filter 271 absorbs this beam and does not transilluminate it in the direction of the image sensor 27. This enables the acquisition of a clear image of fixed-point 31 without noise in the image of fixed-point 31 acquired by the image sensors 27, enhancing the accuracy of the arrangement location information acquired by the image recognition part 281 or 285, and detecting the displacement or locations of the leaf blocks 23B with higher accuracy. Next, Embodiment 5 of the radiotherapy unit 1 of the present invention will be described. The same codes are used for the same structures and same functions as in Embodiments 1 and 2, so detailed descriptions are omitted. The radiotherapy unit 1 of this embodiment detects the displacement and locations in a similar manner to Embodiment 1 or 2. The radiotherapy unit 1 in this embodiment is a modified example of improvement in accuracy of the arrangement location information related to Embodiment 3. FIG. 18 is a block diagram of a detecting element 25 in this embodiment. The irradiation part 26 placed corresponding to each of the leaf blocks 23B irradiates a beam with a wavelength different from the wavelength of beams irradiated by the adjacent irradiation parts 26. The detecting element 25 is provided with an image extraction part 288 electrically connected between the image sensor 27 and the image recognition part 281 (or 285). The image extraction part 288 extracts only the signal of an image based on a beam with a unique wavelength of the irradiation part 26 placed correspondingly from the image of fixed-point 31 acquired by the image sensor 27 placed correspondingly in the same way. Signals of images converted by the image sensor 27 subject to beams irradiated by the adjacent irradiation parts 26 are eliminated by this image extraction part 288. This enables the acquisition of a clear image of fixed-point 31 without pattern images drawn on the adjacent leaf blocks 23B in the image of fixed-point 31 being input into the image recognition part 281 (or 285), thereby enhancing the accuracy of the arrangement location information acquired by the image recognition parts 281 or 285, and detecting the displacement or locations of the leaf blocks 23B with higher accuracy.
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	claims	1. A multileaf collimator, comprising:a plurality of beam-blocking leaves of a first type, each of the plurality of beam-blocking leaves of the first type having a trapezoidal geometry viewed in a longitudinal moving direction with a first lateral side, a second lateral side, a wider end, and a narrower end, the wider end being proximal to a source;a plurality of beam-blocking leaves of a second type, each of the plurality of beam-blocking leaves of the second type having a trapezoidal geometry viewed in the longitudinal moving direction with a first lateral side, a second lateral side, a wider end, and a narrower end, the wider end being distal to the source; andthe plurality of beam-blocking leaves of the first type being alternatingly arranged with the plurality of beam-blocking leaves of the second type side by side. 2. The multileaf collimator of claim 1, whereinfirst lateral sides of the plurality of beam-blocking leaves of the first type align to converge to a first point offset from the source, andsecond lateral sides of the plurality of beam-blocking leaves of the first type align to converge to a second point offset from the source and opposite to the first point. 3. The multileaf collimator of claim 2, whereinwherein a first lateral side of a beam-blocking leaf of the second type is adjacent to a second lateral side of a beam-blocking leaf of the first type, and a second lateral side of a beam-blocking leaf of the second type is adjacent to a first lateral side of a beam-blocking leaf of the first type, andfirst lateral sides of the plurality of beam-blocking leaves of the second type align to converge to the second point, and second lateral sides of the plurality of beam-blocking leaves of the second type align to converge to the first point. 4. The multileaf collimator of claim 3, wherein the first point offsets from the source at a distance substantially equal to a distance that the second point offsets from the source. 5. The multileaf collimator of claim 1, wherein a first lateral side and a second lateral side of each of the plurality of beam-blocking leaves of the first type, and a first lateral side and a second lateral side of each of the plurality of beam-blocking leaves of the second type are substantially flat. 6. The multileaf collimator of claim 5, wherein the first point offsets from the source at a distance substantially equal to a distance that the second point offsets from the source. 7. The multileaf collimator of claim 1, wherein a first lateral side and a second lateral side of each of the plurality of beam-blocking leaves of the first type, and a first lateral side and a second lateral side of each of the plurality of beam-blocking leaves of the second type are substantially flat. 8. A multi-level multileaf collimator (MLC), comprising:a first MLC in a first level distal to a source; anda second MLC in a second level proximal to the source,wherein the second MLC comprises:a plurality of beam-blocking leaves of a first type, each of the plurality of beam-blocking leaves of the first type having a trapezoidal geometry viewed in a longitudinal moving direction with a first lateral side, a second lateral side, a wider end and a narrower end, the wider end being proximal to the source;a plurality of beam-blocking leaves of a second type, each of the plurality of beam-blocking leaves of the second type having a trapezoidal geometry viewed in the longitudinal moving direction with a first lateral side, a second lateral side, a wider end, and a narrower end, the wider end being distal to the source; andthe plurality of beam-blocking leaves of the first type being alternatingly arranged with the plurality of beam-blocking leaves of the second type side by side. 9. The multi-level multileaf collimator of claim 8, whereinfirst lateral sides of the plurality of beam-blocking leaves of the first type align to converge to a first point offset from the source, andsecond lateral sides of the plurality of beam-blocking leaves of the first type align to converge to a second point offset from the source opposite to the first point. 10. The multi-level multileaf collimator of claim 9, whereinwherein a first lateral side of a beam-blocking leaf of the second type is adjacent to a second lateral side of a beam-blocking leaf of the first type, and a second lateral side of a beam-blocking leaf of the second type is adjacent to a first lateral side of a beam-blocking leaf of the first type, andfirst lateral sides of the plurality of beam-blocking leaves of the second type align to converge to the second point, and second lateral sides of the plurality of beam-blocking leaves of the second type align to converge to the first point. 11. The multi-level multileaf collimator of claim 10, wherein the first point offsets from the source at a distance substantially equal to a distance that the second point offsets from the source. 12. The multi-level multileaf collimator of claim 8, wherein a first lateral side and a second lateral side of each of the plurality of beam-blocking leaves of the first type, and a first lateral side and a second lateral side of each of the plurality of beam-blocking leaves of the second type are substantially flat. 13. The multi-level multileaf collimator of claim 8, wherein the first MLC comprises a plurality of beam-blocking leaves, each being longitudinally movable in a direction substantially parallel with the longitudinal moving direction of the plurality of beam-blocking leaves of the first type and the plurality of beam-blocking leaves of the second type of the second MLC. 14. The multi-level multileaf collimator of claim 13, wherein each of the plurality of beam-blocking leaves of the first type and the plurality of beam-blocking leaves of the second type of the second MLC laterally offsets a beam-blocking leaf of the first MLC. 15. An apparatus, comprising:a source of radiation, anda multileaf collimator comprising:a plurality of beam-blocking leaves of a first type, each of the plurality of beam-blocking leaves of the first type having a trapezoidal geometry viewed in a longitudinal moving direction with a first lateral side, a second lateral side, a wider end, and a narrower end, the wider end being proximal the source of radiation;a plurality of beam-blocking leaves of a second type, each of the plurality of beam-blocking leaves of the second type having a trapezoidal geometry viewed in the longitudinal moving direction with a first lateral side, a second lateral side, a wider end, and a narrower end, the wider end being distal to the source of radiation; andthe plurality of beam-blocking leaves of the first type being alternatingly arranged with the plurality of beam-blocking leaves of the second type side by side. 16. The apparatus of claim 15, wherein the source of radiation comprises a source of x-rays, a source of gamma rays, a source of protons, or a source of heavy ions. 17. The apparatus of claim 15, wherein the multileaf collimator comprises a first multileaf collimator in a first level distal to the source of radiation and a second multileaf collimator in a second level proximal to the source of radiation, wherein the plurality of beam-blocking leaves of the first type and the plurality of beam-blocking leaves of the second type are arranged in the second multileaf collimator. 18. The apparatus of claim 17, wherein the first multileaf collimator comprises a plurality of beam-blocking leaves, and each of the plurality of beam-blocking leaves of the first type and the plurality of beam-blocking leaves of the second type of the second multileaf collimator laterally offsets one of the plurality of beam-blocking leaves of the first multileaf collimator. 19. The apparatus of claim 15, whereinfirst lateral sides of the plurality of beam-blocking leaves of the first type align to converge to a first point offset from the source, andsecond lateral sides of the plurality of beam-blocking leaves of the first type align to converge to a second point offset from the source opposite to the first point. 20. The apparatus of claim 19, whereina first lateral side of a beam-blocking leaf of the second type is adjacent to a second lateral side of a beam-blocking leaf of the first type, and a second lateral side of a beam-blocking leaf of the second type is adjacent to a first lateral side of a beam-blocking leaf of the first type, andfirst lateral sides of the plurality of beam-blocking leaves of the second type align to converge to the second point, and second lateral sides of the plurality of beam-blocking leaves of the second type align to converge to the first point.
	summary
062381385	abstract	A method of disposing nuclear waste in underground rock formations (18). The method includes the steps of selecting an land area having a rock formation (18) positioned therebelow of a depth able to prevent radioactive material placed therein from reaching the surface and must be at least a predetermined distance from active water sources and drilling a vertical wellbore (14) from the surface into the underground rock formation (18). A primary horizontal lateral (20) is drilled from the vertical wellbore (14) with the surface of the primary horizontal lateral (20) defined by the underground rock formation (18). A layer of cement (30) is placed within the primary horizontal lateral (20) and a layer of steel (32) is secured within the layer of cement (30). Nuclear waste to be stored within the lateral is placed in a canister (38) and the encapsulated nuclear waste is positioned within the primary horizontal lateral (20). The primary horizontal lateral (20) is then filled with cement (48) to seal the nuclear waste therein. Additional primary horizontal laterals (20) may be drilled from the vertical wellbore (14) and secondary and tertiary horizontal laterals (24, 26) can be drilled from the primary horizontal lateral (20). Additional layers of lead, cement and steel may be used to cover the laterals and shield the rock formation (18) from any radiation leakage. Furthermore, front and end plugs (49, 50) may be positioned at either end of the laterals, retaining the canisters (38) therein and providing added protection from leakage.
	description	This application claims the benefit of Korean Patent Application No. 10-2015-0077487, filed on Jun. 1, 2015, in the Korean Intellectual Property Office, the disclosure of which is incorporated herein in its entirety by reference for all purposes. 1. Field One or more embodiments relate to an X-ray imaging apparatus and a method of operating the X-ray imaging apparatus. 2. Description of the Related Art As the quality of medical care improves with rapid economic growth due to industrialization, there is increasing demand for apparatuses capable of capturing images of objects by using X-rays. Accordingly, there is increasing demand for X-ray imaging apparatuses. X-ray imaging apparatuses for medical use are apparatuses use X-rays having excellent penetration capability to irradiate a targeted region of a human body and capture images of the inside of the body. An X-ray is a form of radiation. If the human body is exposed to radiation, it may damage body tissues and may cause a variety of diseases. Including an X-ray radiation dose reduction device to minimize side effects on a patient due to radiation exposure in X-ray imaging, apparatuses can prevent such problems by minimizing side effects on a patient due to radiation exposure. Including a collimator in X-ray imaging apparatuses for medical can appropriately adjust a range of X-ray radiation by regulating a range of X-rays moving vertically and horizontally. The collimator may typically include irises which may move vertically and horizontally to appropriately adjust a range of X-ray radiation. For example, with some imaging methods, it may be beneficial to regulate the range of radiation of the collimator automatically or manually and to confirm the location of the collimator by using collimator light in order to limit the range of irradiation to a selected region of interest for precise imaging. However, this may be inconvenient and lead to an increased imaging time and inefficient operation motions on the part of users who operate the apparatuses. Provided are X-ray imaging apparatuses including a collimator configured to adjust an irradiation range of the X-ray irradiated from an X-ray source and methods of operating the X-ray imaging apparatuses. According to an aspect of an embodiment, an X-ray imaging apparatus includes: an X-ray source configured to irradiate an X-ray; and a collimator configured to adjust an irradiation range of the X-ray irradiated from the X-ray source, wherein the collimator comprises: a first field size range adjustor comprising a first plurality of blades and a driving power transfer unit configured to transfer driving power to the first plurality of blades; a second field size range adjustor facing the first field size range adjustor and comprising a first plurality of blades; and a connector configured to respectively connect the first plurality of blades of the first field size range adjustor to the first plurality of blades of the second field size range adjustor so as to make the first plurality of blades of the second field size range adjustor move as the plurality of first plurality of blades of the first field size range adjustor move. The driving power transfer unit may include a timing belt. The timing belt may include a plurality of timing belts, wherein the plurality of timing belts includes a first timing belt and a second timing belt, the first plurality of blades of the first field size range adjustor includes a first blade and a second blade, a third blade and a fourth blade, wherein the first blade and the second blade of the first plurality of blades of the first field size range adjustor face each other and are fixed to the first timing belt; and the third and fourth blade of the first plurality of blades of the first field size range adjustor face each other and are fixed to the second timing belt. The X-ray imaging apparatus may further include: a plurality of driving motors configured to transfer driving power to the timing belts including a first driving motor and a second driving motor, wherein the first driving motor is configured to transfer driving power to the first timing belt and the second driving motor is configured to transfer driving power to the second timing belt. The X-ray imaging apparatus may further include: a controller configured to transfer a driving signal to the first driving motor and the second driving motor, wherein the controller transfers different control signals to the first driving motor and the second driving motor such that the first and second blade of the first plurality of blades of the first field range adjustor blades move independently from the third and fourth blade of the first plurality of blades of the first field range adjustor. The X-ray imaging apparatus may further include: a first slide support unit configured to allow the first and second blade of the first plurality of blades of the first field range adjustor, the first slide support unit controlling a movement path of the first and second blade of the first plurality of blades of the first field range adjustor; and a second slide support unit configured to allow third and fourth blade of the first plurality of blades of the first field range adjustor to slide, the second slide support unit controlling the third and fourth blade of the first plurality of blades of the first field range adjustor. The timing belt may include a plurality of timing belts, wherein the first and second blade of the first plurality of blades of the first field range adjustor which face each other and the third and fourth blade of the first plurality of blades of the first field range adjustor which face each other are respectively fixed to the plurality of timing belts. The X-ray imaging apparatus may further include: a plurality of driving motors configured to transfer driving power to the timing belts; and wherein the plurality of driving motors are further configured to respectively transfer driving power to the plurality of timing belts. The X-ray imaging apparatus may further include: a controller configured to transfer the driving signal to the plurality of driving motors, wherein the controller transfers different control signals to the plurality of driving motors such that the first, second, third, and fourth blades of the first plurality of blades of the first field range adjustor to move independently from each other. The X-ray imaging apparatus may further include: a first slide support unit configured to allow the first and second blade of the first plurality of blades of the first field range adjustor to slide, the first slide support unit controlling a movement path of the first and second blade of the first plurality of blades of the first field range adjustor; and a second slide support unit configured to allow the third and fourth blade of the first plurality of blades of the first field range adjustor to slide, the second slide support unit controlling a movement path of the third and fourth blade of the first plurality of blades of the first field range adjustor. The connector may include a connecting link arranged to revolve around a hinge unit, and a linking unit configured to respectively connect the first blades and the second blades by using the connecting link. The linking unit may further include: a first long hole extending along a lengthwise direction of the connecting link in the surface of the connecting link; a first slider configured to be fixed to the first blade and to slide along a lengthwise direction of the first long hole; a second long hole extending along the lengthwise direction of the connecting link in the surface of the connecting link; and a second slider configured to be fixed to the second blade and to slide along a lengthwise direction of the second long hole. A movement ratio of the first blade to the second blade which is connected to the first blade may be the same as a distance ratio of a distance from the hinge unit to the first slider to a distance from the hinge unit to the second slider. According to another aspect of an embodiment, a method of operating the X-ray imaging apparatus described above includes: inputting an adjustment signal corresponding to a first field size range and an adjustment signal corresponding to a second field size range; moving the plurality of first blades; and moving the plurality of second blades in synchronization with the plurality of the first blades. In the method, the plurality of the first blades may include a first, second, third, and fourth blade of the first plurality of blades of the first field range adjustor, and the first plurality fo blades of the second field range adjustor may include a first, second, third, and fourth blade, wherein the method may further include: generating a driving signal for a first driving motor and a second driving motor upon receiving the adjustment signal; generating driving power by the first driving motor and the second driving motor upon receiving the driving signal; and transferring the driving power generated by the first driving motor to the first and second blade of the first plurality of blades of the first field range adjustor and the driving power generated by the second driving motor to the third and fourth blade of the first plurality of blades of the first field range adjustor, respectively. The method may further include providing independent driving signals to the first driving motor and the second driving motor; and moving the first and second blade of the first plurality of blades of the first field range adjustor independently from the third and fourth blade of the first plurality of blades of the first field range adjustor upon receiving the driving signal. In the method, the first plurality of the blades of the first field range adjustor may include a 1st blade, a 2nd blade, a 3rd blade, and a 4th blade, and the first plurality of blades from the second field range adjustor may include a 1st blade, a 2nd blade, a 3rd blade, and a 4th blade; and wherein the method may further include: generating a driving signal for a plurality of driving motors upon receiving the adjustment signal; generating driving power at the plurality of driving motors upon receiving the driving signal; and transferring the driving power generated at the plurality of driving motors to each of the first, second, third, and fourth blades of the first plurality of blades of the first field range adjustor, respectively. In the method, independent driving signals may be respectively provided to the plurality of driving motors; and the first, second, third, and fourth blades of the first plurality of blades of the first field range adjustor are moved independently from each other. The attached drawings for illustrating embodiments of the present disclosure are referred to in order to gain a sufficient understanding of the present disclosure, the merits thereof, and the objectives accomplished by the implementation of the present disclosure. In this regard, the present embodiments may have different forms and should not be construed as being limited to the descriptions set forth herein. Rather, these embodiments are provided so that this disclosure will be thorough and complete and will fully convey the concept of the present embodiments to one of ordinary skill in the art, and the present invention will only be defined by the appended claims. Hereinafter, the terms used in the specification will be briefly described, and then the present disclosure will be described in detail. The terms used in this specification are those general terms currently widely used in the art in consideration of functions regarding the present disclosure, but the terms may vary according to the intention of those of ordinary skill in the art, precedents, or new technology in the art. Also, some terms may be arbitrarily selected by the applicant, and in this case, the meaning of the selected terms will be described in detail in the detailed description of the present specification. Thus, the terms used in the specification should be understood not as simple names but based on the meaning of the terms and the overall description of the invention. Throughout the specification, an “image” may denote multi-dimensional data composed of discrete image elements (for example, pixels in a two-dimensional image and voxels in a three-dimensional image). For example, an image may be a medical image of an object acquired by X-ray imaging apparatus, a computed tomography (CT) apparatus, a magnetic resonance imaging (MRI) apparatus, an ultrasound diagnosis apparatus, or another medical imaging apparatus. In addition, an “object” may be a human, an animal, or a part of a human or animal. For example, the object may include an organ (for example, the liver, the heart, the womb, the brain, breasts, or the abdomen), blood vessels, or a combination thereof. The object may be a phantom. The phantom denotes a material having a volume, a density, and an effective atomic number that are approximately the same as those of a living organism. For example, the phantom may be a spherical phantom having similar properties to those of the human body. Throughout the specification, a “user” may be, but is not limited to, a medical expert, for example, a medical doctor, a nurse, a medical laboratory technologist, or a medical imaging expert, or a technician who repairs medical apparatuses. X-ray imaging apparatus is a medical imaging apparatus that acquires images of internal structures of an object by transmitting X-ray through the human body. The X-ray imaging apparatus may acquire medical images of an object more simply within a shorter time than other medical imaging apparatuses including an MRI apparatus and a CT apparatus. Therefore, the X-ray imaging apparatus is widely used in simple chest imaging, simple abdomen imaging, simple skeleton imaging, simple nasal sinuses imaging, simple neck soft tissue imaging, and breast imaging. FIG. 1 is a view illustrating a structure of a X-ray imaging apparatus according to the related art. The X-ray imaging apparatus 10 shown in FIG. 1 may be a fixed-type X-ray imaging apparatus or a mobile X-ray imaging apparatus. Referring to FIG. 1, the X-ray imaging apparatus 10 includes a workstation 11, X-ray irradiation unit 12, a high voltage generator 121, and an X-ray detector 13. The workstation 11 includes an input unit 112 through which a user may input commands for manipulating the X-ray imaging apparatus 10 including X-ray irradiation, and a controller 113 controlling overall operations of the X-ray imaging apparatus 10. The high voltage generator 121 generates a high voltage for generating X-rays, and applies the high voltage to X-ray source 122. The X-ray irradiation unit 12 includes the X-ray source 122 receiving the high voltage applied from the high voltage generator 121 to generate and irradiate the X-ray, and a collimator 300 for guiding a path of the X-ray irradiated from the X-ray source 122. The collimator 300 can include a first field size range adjustor having a plurality of blades and a driving power transfer unit to transfer driving power to the plurality of blades, a second field range adjustor facing the first field size range adjustor having a plurality of blades, and a connector connecting the plurality of blades of the first field size range adjustor and the second field size range adjustor, so that the second plurality of blades moves with the first plurality of blades. The collimator 300 will be described in greater detail in FIGS. 4-13. The X-ray detector 13 detects X-ray that is radiated from the X-ray irradiation unit 12 and has been transmitted through an object. Also, the X-ray imaging apparatus 10 may further include an operating unit 14 including a sound output unit 141 outputting sound representing information relating to imaging operation such as the X-ray irradiation under a control of the controller 113. The workstation 11, the X-ray irradiation unit 12, the high voltage generator 121, and the X-ray detector 13 may be connected to each other via wires or wirelessly. If they are connected to each other wirelessly, a device (not shown) for synchronizing clocks with each other may be further included. The input unit 112 may include a keyboard, a mouse, a touch screen, a voice recognizer, a fingerprint recognizer, an iris recognizer, and the like well known in the art. The user may input a command for irradiating the X-ray via the input unit 112, and to do this, the input unit 112 may include a switch for inputting the command. The switch may be configured so that an irradiation command for irradiating the X-ray may be input only when the switch is pushed twice. That is, when the user pushes the switch, a prepare command for performing a pre-heating operation for X-ray irradiation may be input through the switch, and then, when the user pushes the switch once more, the irradiation command for irradiating the X-ray may be substantially input through the switch. When the user manipulates the switch as described above, the input unit 112 generates signals corresponding to the commands input through the switch manipulation, that is, a prepare signal and an irradiation signal, and outputs the generated signals to the high voltage generator 121 generating a high voltage for generating the X-ray. When the high voltage generator 121 receives the prepare signal output from the input unit 112, the high voltage generator 121 starts a pre-heating operation, and when the pre-heating is finished, the high voltage generator 121 outputs a ready signal to the controller 121. In addition, the X-ray detector 13 also needs to prepare for detecting the X-ray, and thus, when the high voltage generator 121 receives the prepare signal output from the input unit 112, the high voltage generator 121 outputs a prepare signal to the X-ray detector 13 at the same time of performing the pre-heating operation, so that the X-ray detector 13 may prepare for detecting the X-ray transmitted through the object. The X-ray detector 13 prepares for detecting the X-ray when receiving the prepare signal, and when the preparing for the detection is finished, the X-ray detector 130 outputs a ready signal to the high voltage generator 121 and the controller 113. When the pre-heating operation of the high voltage generator 121 is finished, the X-ray detector 13 is ready for the detecting the X-ray, and the irradiation signal is output from the input unit 112 to the high voltage generator 121, the high voltage generator 121 generates and applies the high voltage to the X-ray source 122, and the X-ray source 122 irradiates the X-ray. When the irradiation signal is output from the input unit 112, the controller 113 may output a sound output signal to the sound output unit 141 so that the sound output unit 141 outputs predetermined sound and the object may recognize the irradiation of X-ray. Also, the sound output unit 141 may output sound representing other information relating to the imaging, in addition to the X-ray irradiation. In FIG. 1, the sound output unit 141 is included in the operating unit 14; however, the embodiments of the present disclosure are not limited thereto, and the sound output unit 141 may be located at a different location from the operating unit 14. For example, the sound output unit 141 may be included in the workstation 11, or may be located on a wall surface of an examination room in which the X-ray imaging of the object is performed. The controller 113 controls locations of the X-ray irradiation unit 12 and the X-ray detector 13, an imaging timing, and imaging conditions according to imaging conditions set by the user. In more detail, the controller 113 controls the high voltage generator 121 and the X-ray detector 13 according to the command input via the input unit 112 so as to control an irradiation timing of the X-ray, an intensity of the X-ray, and an irradiation region of the X-ray. Also, the controller 113 adjusts the location of the X-ray detector 13 according to a predetermined imaging condition, and controls an operation timing of the X-ray detector 13. In addition, the controller 113 generates a medical image of the object by using image data transmitted from the X-ray detector 13. In detail, the controllers 112 may receive the image data from the X-ray detector 13, and then, generate the medical image of the object by removing noise from the image data and adjusting a dynamic range and interleaving of the image data. The X-ray imaging apparatus 10 shown in FIG. 1 may further include an output unit (not shown) for outputting the medical image generated by the controller 113. The output unit may output information that is necessary for the user to manipulate the X-ray imaging apparatus 10, for example, a user interface (UI), user information, or object information. The output unit may include a printer, a cathode ray tube (CRT) display, a liquid crystal display (LCD), a plasma display panel (PDP), an organic light emitting diode (OLED) display, a field emission display (FED), a light emitting diode (LED) display, a vacuum fluorescent display (VFD), a digital light processing (DLP) display, a primary flight display (PFD), a three-dimensional (3D) display, a transparent display, and other various output devices well known in the art. The workstation 11 shown in FIG. 1 may further include a communicator (not shown) that may be connected to a server 162, a medical apparatus 164, and a portable terminal 166 via a network 16. The communication unit may be connected to the network 16 via wires or wirelessly to communicate with the external server 162, the external medical apparatus 164, or the external portable terminal 166. The communicator may transmit or receive data related to diagnosis of the object via the network 16, and may also transmit or receive medical images captured by the medical apparatus 164, for example, a CT apparatus, an MRI apparatus, or X-ray imaging apparatus. Moreover, the communicator may receive a medical history or treatment schedule of an object (e.g., a patient) from the server 162 to diagnose a disease of the object. Also, the communicator may perform data communication with the portable terminal 166 such as a mobile phone, a personal digital assistant (PDA), or a laptop computer of a medical doctor or a client, as well as the server 162 or the medical apparatus 164 in a hospital. The communicator may include one or more elements enabling communication with external apparatuses. For example, the communicator may include a local area communication module, a wired communication module, and a wireless communication module. The local area communication module refers to a module for performing local area communication with an apparatus located within a predetermined distance. Examples of local area communication techniques according to an embodiment may include, but are not limited to, wireless LAN, Wi-Fi, Bluetooth, ZigBee, Wi-Fi Direct (WFD), ultra wideband (UWB), infrared data association (IrDA), Bluetooth low energy (BLE), and near field communication (NFC). The wired communication module is a module for communicating by using an electric signal or an optical signal, and the wired communication technology may be wired communication technology using a pair cable, a coaxial cable, or an optical fiber cable, and a wired communication technology that is well known in the art. The wireless communication module may transmit/receive a wireless signal to/from at least one of a base, an external device, and a server in a mobile communication network. Here, the wireless signal may be a voice call signal, a video call signal, or various types of data according to text/multimedia messages transmission. The X-ray imaging apparatus 10 shown in FIG. 1 may include a plurality of digital signal processors (DSPs), an ultra-small calculator, and a processing circuit for special purposes (for example, high speed analog/digital (A/D) conversion, high speed Fourier transformation, and an array process). In addition, the communication between the workstation 11 and the X-ray irradiation unit 12, the workstation 11 and the high voltage generator 121, and the workstation 11 and the X-ray detector 13 may use a high speed digital interface, such as low voltage differential signalling (LVDS), asynchronous serial communication, such as universal asynchronous receiver transmitter (UART), synchronous serial communication, or a low latency network protocol, such as a controller area network (CAN), and other various communication methods that are well known in the art may be used.) FIG. 2 is a perspective view illustrating a fixed-type X-ray imaging apparatus, according to an embodiment. Referring to FIG. 2, the fixed type X-ray imaging apparatus 20 includes an operating unit 14 providing a user with an interface for manipulating the X-ray imaging apparatus 20, X-ray irradiation unit 12 radiating X-ray to an object, a X-ray detector 13 detecting X-ray that has passed through the object, motors 211, 212, and 213 providing a driving power to transport the X-ray irradiation unit 12, a guide rail 22, a moving carriage 23, and a post frame 24. The guide rail 22, the moving carriage 23, and the post frame 24 are formed to transport the X-ray irradiation unit 12 by using the driving power of the motors 211, 212, and 213. The guide rail 22 includes a first guide rail 221 and a second guide rail 222 that are provided to form a predetermined angle with respect to each other. The first guide rail 221 and the second guide rail 222 may be substantially perpendicular to each other, respectively extend in directions crossing each other at substantially 90°. The first guide rail 221 can be provided on the ceiling of an examination room in which the X-ray imaging apparatus 20 is disposed. The second guide rail 222 can be located under the first guide rail 221, and mounted so as to slide along the first guide rail 221. A roller (not shown) that may move along the first guide rail 221 may be provided on the first guide rail 221. The second guide rail 222 is connected to the roller to move along the first guide rail 221. A first direction D1 is defined as a direction in which the first guide rail 221 extends, and a second direction D2 is defined as a direction in which the second guide rail 222 extends. Therefore, the first direction D1 and the second direction D2 cross each other at 90°, and may be parallel to the ceiling of the examination room. The moving carriage 23 is disposed under the second guide rail 222 so as to move along the second guide rail 222. A roller (not shown) moving along the second guide rail 222 may be provided on the moving carriage 23. Therefore, the moving carriage 23 may move in the first direction D1 together with the second guide rail 222, and may move in the second direction D2 along the second guide rail 222. The post frame 24 is fixed on the moving carriage 23 and located under the moving carriage 23. The post frame 24 may include a plurality of posts 241, 242, 243, 244, and 245. The plurality of posts 241, 242, 243, 244, and 245 are telescopically connected to each other, and thus, the post frame 24 may have a length that is adjustable in a vertical direction of the examination room while in a state of being fixed to the moving carriage 23. A third direction D3 is defined as a direction in which the length of the post frame 24 increases or decreases. Therefore, the third direction D3 may be perpendicular to the first direction D1 and the second direction D2. The X-ray irradiation unit 12 may include the X-ray source 122 and the collimator 300 which regulates the radiation range of X-rays generated and irradiated by the X-ray source 122. The X-ray source 122 includes an X-ray tube that may be realized as a vacuum tube diode including a cathode and an anode. An inside of the X-ray tube is set as a high vacuum state of about 10 mmHg, and a filament of the anode is heated to a high temperature to generate thermal electrons. The filament may be a tungsten filament, and a voltage of about 10V and a current of about 3 to 5 A may be applied to an electric wire connected to the filament to heat the filament. In addition, when a high voltage of about 10 to about 300 kVp is applied between the cathode and the anode, the thermal electrons are accelerated to collide with a target material of the cathode, and then, X-ray is generated. The X-ray is radiated outside via a window, and the window may be formed of a beryllium thin film. In this case, most of the energy of the electrons colliding with the target material is consumed as heat, and remaining energy is converted into the X-ray. The cathode is mainly formed of copper, and the target material is disposed opposite to the anode. The target material may be a high resistive material such as chromium (Cr), iron (Fe), cobalt (Co), nickel (Ni), tungsten (W), or molybdenum (Mo). The target material may be rotated by a rotating field. When the target material is rotated, an electron impact area is increased, and a heat accumulation rate per unit area may be increased to be at least ten times greater than that of a case where the target material is fixed. The voltage applied between the cathode and the anode of the X-ray tube is referred to as a tube voltage, and the tube voltage is applied from the high voltage generator 121 and a magnitude of the tube voltage may be expressed by a crest value (kVp). When the tube voltage increases, a velocity of the thermal electrons increases, and accordingly, an energy of the X-ray (energy of photon) that is generated when the thermal electrons collide with the target material is increased. The current flowing in the X-ray tube is referred to as a tube current that may be expressed as an average value (mA). When the tube current increases, the number of thermal electrons emitted from the filament is increased, and accordingly, the X-ray dose (the number of X-ray photons) generated when the thermal electrons collide with the target material is increased. Therefore, the energy of the X-ray may be adjusted according to the tube voltage, and the intensity of the X-ray or the X-ray dose may be adjusted according to the tube current and the X-ray exposure time. The high voltage generator 121 may be included in the X-ray source 122, but is not limited thereto, and may be included somewhere else in the X-ray imaging apparatus 20. The X-ray detector 13 detects X-rays that have passed through the object, and may be configured either as a table-type 29 X-ray detector 13 or as a stand-type 28 X-ray detector 13. The X-ray detector 13 may be implemented by using a thin film transistor (TFT) or a charge coupled device (CCD). A rotating joint 25 is disposed between the X-ray irradiation unit 12 and the post frame 24. The rotating joint 25 allows the X-ray irradiation unit 12 to be coupled to the post frame 24, and supports a load applied to the X-ray irradiation unit 12. The X-ray irradiation unit 12 connected to the rotating joint 25 may rotate on a plane that is perpendicular to the third direction D3. In this case, a rotating direction of the X-ray irradiation unit 12 may be defined as a fourth direction D4. Also, the X-ray irradiation unit 12 may be configured to be rotatable on a plane perpendicular to the ceiling of the examination room. Therefore, the X-ray irradiation unit 12 may rotate in a fifth direction D5 that is a rotating direction about an axis that is parallel with the first direction D1 or the second direction D2, with respect to the rotating joint 25. The motors 211, 212, and 213 may be provided to move the X-ray irradiation unit 12 in the first, second, and third directions D1, D2, and D3. The motors 211, 212, and 213 may be electrically driven, and the motors 211, 212, and 213 may respectively include an encoder. The motors 211, 212, and 213 may be disposed at various locations in consideration of design convenience. For example, the first motor 211, moving the second guide rail 222 in the first direction D1, may be disposed around the first guide rail 221, the second motor 212, moving the moving carriage 230 in the second direction D2, may be disposed around the second guide rail 222, and the third motor 213, increasing or reducing the length of the post frame 240 in the third direction D3, may be disposed in the moving carriage 23. In another example, the motors 211, 212, and 213 may be connected to a driving power transfer unit (not shown) so as to linearly move the X-ray irradiation unit 12 in the first, second, and third directions D1, D2, and D3. The driving power transfer unit may be a combination of a belt and a pulley, a combination of a chain and a sprocket, or a shaft, which are generally used. In another example, motors (not shown) may be disposed between the rotating joint 25 and the post frame 24 and between the rotating joint 25 and the X-ray irradiation unit 12 in order to rotate the X-ray irradiation unit 12 in the fourth and fifth directions D4 and D5. On one side of the X-ray irradiation unit 12, an operating unit 14 is included which provides an interface that allows entering of various input information and controlling of each device. Although FIG. 2 shows the fixed type X-ray imaging apparatus 20 connected to the ceiling of the examination room, the fixed type X-ray imaging apparatus 20 is merely an example for convenience of comprehension. That is, X-ray imaging apparatuses according to embodiments of the present disclosure may include X-ray imaging apparatuses having various other structures such as, for example, a C-arm-type X-ray imaging apparatus and an angiography X-ray imaging apparatus, in addition to the fixed type X-ray imaging apparatus 20 of FIG. 2. FIG. 3 is a view illustrating the mobile X-ray imaging apparatus capable of capturing X-ray images without being affected by an imaging location, according to an embodiment. An X-ray apparatus 30 shown in FIG. 3 may include a moving unit 37 including a wheel for facilitating the movement of the X-ray apparatus 30; a main unit including an input unit 142 receiving commands for operating the X-ray apparatus 30, a high voltage generator 121 generating high voltage applied to the X-ray source 122, a sound output unit 141 outputting sounds indicating imaging-related information such as X-ray irradiation, and a controller 15 controlling overall operation of the X-ray apparatus 30; the X-ray irradiation unit 12 including the X-ray source 122 generating X-rays and the collimator 300 for guiding a path of the X-rays irradiated from the X-ray source 122; and the X-ray detector 13 which detects X-rays that are radiated from the X-ray irradiation unit 12 and have penetrated an object. The input unit 142 receives some input from the user. The input unit 142 may include a keyboard, a mouse, a touch screen, a voice recognizer, a fingerprint recognizer, an iris recognizer, and the like well known in the art. The user may input a command for irradiating the X-ray via the input unit 142, and to do this, the input unit 142 may include a switch for inputting the command. The switch may be configured so that an irradiation command for irradiating the X-ray may be input only when the switch is pushed twice. That is, when the user pushes the switch, a prepare command for performing a pre-heating operation for X-ray irradiation may be input through the switch, and then, when the user pushes the switch once more, the irradiation command for irradiating the X-ray may be substantially input through the switch. When the user manipulates the switch as described above, the input unit 142 generates signals corresponding to the commands input through the switch manipulation, that is, a prepare signal and an irradiation signal, and outputs the generated signals to the high voltage generator 121 generating a high voltage for generating the X-ray. When the high voltage generator 121 receives the prepare signal output from the input unit 142, the high voltage generator 121 starts a pre-heating operation, and when the pre-heating is finished, the high voltage generator 121 outputs a ready signal to the controller 15. In addition, the X-ray detector 13 also needs to prepare for detecting the X-ray, and thus, when the high voltage generator 121 receives the prepare signal output from the input unit 142, the high voltage generator 121 outputs a prepare signal to the X-ray detector 13 at the same time of performing the pre-heating operation, so that the X-ray detector 13 may prepare for detecting the X-ray transmitted through the object. The X-ray detector 13 prepares for detecting the X-ray when receiving the prepare signal, and when the preparing for the detection is finished, the X-ray detector 130 outputs a ready signal to the high voltage generator 121 and the controller 15. When the pre-heating operation of the high voltage generator 121 is finished, the X-ray detector 13 is ready for the detecting the X-ray, and the irradiation signal is output from the input unit 142 to the high voltage generator 121, the high voltage generator 121 generates and applies the high voltage to the X-ray source 122, and the X-ray source 122 irradiates the X-ray. When the irradiation signal is output from the input unit 142, the controller 15 may output a sound output signal to the sound output unit 141 so that the sound output unit 141 outputs predetermined sound and the object may recognize the irradiation of X-ray. Also, the sound output unit 141 may output sound representing other information relating to the imaging, in addition to the X-ray irradiation. In FIG. 3, the sound output unit 141 is included in the main unit 31; however, embodiments are not limited thereto. For example, the sound output unit 141 may be located where the mobile X-ray apparatus 30 is located (e.g., on a wall of a hospital room(. The controller 15 controls locations of the X-ray irradiation unit 12 and the X-ray detector 13, an imaging timing, and imaging conditions according to imaging conditions set by the user. In addition, the controller 15 generates a medical image of the object by using image data transmitted from the X-ray detector 13. In detail, the controllers 113 and 15 may receive the image data from the X-ray detector 13, and then, generate the medical image of the object by removing noise from the image data and adjusting a dynamic range and interleaving of the image data. The main unit 31 of the X-ray apparatus 30 shown in FIG. 3 may further include an output unit (not shown) which outputs a medical image generated by the controller 15. The output unit may output information that is necessary for the user to manipulate the X-ray imaging apparatus 30, for example, a user interface (UI), user information, or object information. FIG. 4 is schematic view of X-ray irradiation unit 12, according to an embodiment. FIG. 5 is a perspective view illustrating a collimator, according to an embodiment. FIG. 6A is a partially-cut perspective view illustrating a first iris unit and a first blocking unit included in the collimator as shown in FIG. 5, according to an embodiment. FIG. 6B is a partial plan view illustrating the first iris unit and the first blocking unit included in the collimator as shown in FIG. 5, according to an embodiment. Referring to FIGS. 4 through 6B, the collimator 300 may include a housing 350 which forms a certain space, a first field size range adjustor for controlling the first field size range P, a second field size range adjustor 500 which is arranged to face the first field size range adjustor 400 so as to control the second field size range T, and a connector 600 arranged between the first field size range adjustor 400 and the second field size range adjustor 500. The first field size range adjustor 400 is an iris unit which is arranged at a bottom end of the housing 350 and may include a first iris unit 410 which adjusts the field size range along the X-axis and a second iris unit 420 which adjusts the field size range along the Z-axis. The a rear view of the first iris unit 410 is shown separated from the remainder of the collimeter 300 in FIG. 6A. The first iris unit 410 may include a blade 411 moving along the X-axis, a blade 412, a first slide support unit 413 which supports the 1st blade 411 and the blade 412, a first driving motor 414-1 which generates driving power to move the blade 411 and the blade 412, a first driving power transfer unit 415 which may transfer the driving power generated by the first driving motor 414-1 to the blade 411 and the blade 412. In various embodiments, the blade 411 and the blade 412 may be formed in the form of capital letter “L” and may be arranged to be spaced apart and parallel from each other with a gap therebetween. The blade 411 and the blade 412 may be movable along the X-axis and an X-ray may pass through a void formed between the blade 411 and the blade 412. In some embodiments, the blade 411 and the blade 412 may include a material such as lead, bismuth, silver, or tungsten having a property of absorbing X-rays irradiated by the X-ray source 122. Accordingly, it is possible to reduce an external leakage of X-rays which are not able to pass through the collimator 300. However, the present embodiment is not limited thereto, and it is possible to reduce an external leakage of X-rays which are not able to pass through the collimator 300, by providing a special coating film, which includes a material having a property of absorbing X-rays, on one side of the blade 411 and the blade 412. The first slide support unit 413 is a support member to control the blade 411 and the blade 412, and may be arranged so as to slide with the blade 411 and the blade 412. According to one embodiment, the first slide support unit 413 may include a slide support bar 413-1 extending along the X-axis and the first slide member and the second slide member 413-2 and 413-3 that may slide along the slide support bar 413-1 in the direction of the X-axis. The first slide member and the second slide member 413-2 and 413-3 may be connected to slide over the slide support bar 413-1, and may be arranged while being fixed to the blade 411 and the blade 412. Thus, the first slide member and the second slide member 413-2 and 413-3 may move the blade 411 and the blade 412 along the direction of extension of the slide support bar 413-1 (i.e., in the direction of the X-axis(. The first driving motor 414-1 is a driving member capable of generating driving power to move the blade 411 and the blade 412, and an output axis 414-11 of the first driving motor 414-1 may be connected to the first driving power transfer unit 415. The first first driving power transfer unit 415 is a driving power transfer member that may transfer the driving power generated at the first driving motor 414-1 to the blade 411 and the blade 412. In various embodiments, the first driving power transfer unit 415 may include a timing belt 415-11 which is supported to revolve around an output axis 414-11 and move according to a rotation of the output axis 414-11 and a support axis 415-12 which is arranged to face the output axis 414-11 and may support the timing belt 415-11. The timing belt 415-11 may be arranged to revolve around the output axis 414-11 and the support axis 415-12 while being supported by the output axis 414-11 and the support axis 415-12. In this case, the first connecting member 416-1 and the second connecting member 416-2 may be arranged to be fixed to the timing belt 415-11 which moves in different directions along the X-axis. As a result, a first connecting member 416-1 and a second connecting member 416-2 may be moved in different directions along the path of the timing belt 415-11, and the blade 411 and the blade 412, which are arranged to be fixed to the first connecting member 416-1 and the second connecting member 416-2, also may be moved in different directions along the path of the timing belt 415-11. A second iris unit 420 may include a blade 421 moving along the Z-axis, a blade 422, a second slide support unit 423 supporting the blade 421 and the blade 422, a second driving motor 424-1 generating driving power to move the blade 421 and the blade 422, a second driving power transfer unit 425 to transfer driving power generated from the second driving motor 424-1 to the blade 421 and the blade 422. The blade 421 and the blade 422 may be formed in the form of capital letter “L” and may be arranged to be spaced apart with a gap. The blade 421 and the blade 422 may be moveable along the Z-axis and X-rays may pass through a void formed between the blade 421 and the blade 422. The blade 421 and the blade 422 may be arranged on the upper portion of the blade 411 and the blade 412. As a result, X-rays radiating from the X-ray source 122 may be irradiated toward an object by passing through an overlapped area of a void formed between the blade 411 and the blade 412 and a void formed between the blade 421 and the blade 422. As related information about a second slide support unit 423, a second driving motor 424-1 and a second driving power transfer unit 425 included in a second iris unit 420 is virtually the same as that about the first slide support unit 413, the first driving motor 414-1 and the first driving power transfer unit 415 included in the first iris unit 410, redundant descriptions will be omitted herein. The second field size range adjustor 500 is an iris unit which is arranged at a top end of the housing 350 and may include a first blocking unit 510 which adjusts the field size range along the X-axis and a second blocking unit 520 which adjusts the field size range along the Z-axis. The first blocking unit 510 may include the 1st blade 511 and the 2nd blade 512 moveable along the X-axis, and the second blocking unit 520 may include the blade 521 and the blade 522 moving along the Z-axis. The blades 511, 512, 521 and 522 may be plate-like members inclined at angles to an optical axis, and may include a material such as lead, bismuth, silver, or tungsten having a property of absorbing X-rays irradiated by the X-ray source 122. In various embodiments, as shown in FIG. 6, the blade 511 may be formed as a fish bone-type blade, and may include a core unit 511-1 which extends at an angle from the optical axis and a plurality of frame units 511-2 which extend perpendicularly to the direction in which the core unit 511-1 extends. As a result, in the case that the blades 511, 512, 521 and 522 are arranged to be close to one another, the plurality of frame units 511-2 included in the blades 511, 512, 521 and 522 may be arranged to interleave with one another. As a result, the blades 511, 512, 521 and 522 may be moved close to one another without interfering with one another. X-rays may pass through a void formed between the blade 511 and the blade 512 along the X-axis, and may pass through a void formed between the blade 521 and the blade 522 along the Z-axis. Accordingly, in the case that the blade 511 and the blade 512 move closer to each other along the X-axis, a void through which the X-rays pass along the X-axis may become narrow, and in the case that the blade 511 and the blade 512 move farther away from each other along the X-axis, a void through which the X-rays pass along the X-axis may become wide. In some embodiments, in the case that the blade 521 and the blade 522 move closer to each other along the Z-axis, a void through which the X-rays pass along the Z-axis may become narrow, and in the case that the blade 521 and the blade 522 move farther from each other along the Z-axis, a void through which the X-rays pass along the Z-axis may become wide. As previously described in detail, as the blades 511, 512, 521 and 522 may be moved close to one another without interfering with one another, X-rays radiating from the X-ray source 122 may be irradiated toward an object by passing through an overlapped area of a void formed between the blade 511 and the blade 512 and a void formed between the blade 521 and the blade 522. A connector 600 is a connecting member which connects the blades 411, 412, 421 and 422 to the blades 511, 512, 521 and 522, respectively. In various embodiments, the connector 600 may include a linking unit 621 with a connecting unit 620 that may revolve around a hinge unit 610. The linking unit 621 has a linking structure to connect the first blade 410 and the second blade 420 by using the connecting unit 620. As previously described in detail, the connecting unit 620 may revolve around the hinge unit 610 while the first blade 410 and the second blade 420 connected to the connecting unit 620 may move linearly, according to an embodiment. As a result, the linking unit 621 may include the connecting unit 620, a first slide unit 630 which has a link joint slide along the lengthwise direction of the connecting unit 620, and a second slide unit 640. The connecting unit 620 is a linear connecting member formed to be extended in one direction, and may be arranged to revolve around the hinge unit 610 which is provided at one end. In various embodiments, the connecting unit 620 may be revolved as the blades 411, 412, 421 and 422 move, and as the connecting unit 620 revolves, the blades 511, 512, 521 and 522 which are arranged to be fixed to a first slider 632, to be described in detail hereinafter, may move in the direction of the axis X or the axis Z. The first slide unit 630 may include a first long hole 631 which extends along the direction of extension of the connecting unit 620 and a first slider 632 which slides while being inserted into the first long hole 631. The first long hole 631 is a slide guide member which is able to control a movement direction of the first slider 632, and may be arranged at either end of the connecting unit 620. The first slider 632 is a slide member which may slide along the lengthwise direction of the first long hole 631, while being inserted into the first long hole 631. The first slider 632 may be arranged to be fixed to each of the blades 511, 512, 521 and 522, respectively, As a result, the blades 511, 512, 521 and 522 may move together as the first slider 632 slides. The second slide unit 640 may include a second long hole 641 which extends along the direction of extension of the connecting unit 620 and a second slider 642 which slides while being inserted into the second long hole 641. The second long hole 641 is a slide guide member which is able to control a movement direction of the second slide unit 642, and may be arranged between the first long hole 631 and one end of the connecting unit 620 where the hinge unit 610 is not provided. The second slider 642 is a slide member which may slide along the lengthwise direction of the second long hole 641, while being inserted into the second long hole 641. The second slider 642 may be arranged to be fixed to each of the blades 411, 412, 421 and 422, respectively. In this case, a joint 643 may be arranged between the second slide unit 642 and the blades 411, 412, 421 and 422. The joint 643 may be fixed to the second slide unit 642 and the blades 411, 412, 421 and 422. As a result, the second slide unit 642 may be fixed to the blades 411, 412, 421 and 422. However, the present embodiment is not limited thereto, and the second slide unit 642 may be arranged to be fixed directly to each of the blades 411, 412, 421 and 422, respectively. As a result, the second slide unit 642 may slide along the lengthwise direction of the second long hole 641 as the blades 411, 412, 421 and 422 move. FIG. 7 is a block diagram illustrating a structure of the collimator, according to an embodiment. FIG. 8 is a partial front view illustrating the first iris unit which shows a movement status of the blades 411, 412, according to an embodiment. FIG. 9 is a partial plan view illustrating the first blocking unit which shows a movement status of the 1st and 2nd blades 511 and 512, according to an embodiment. Referring to FIG. 7, the blade 411 and the blade 412 included in the first iris unit 410 may be moved by receiving driving power generated by the first driving motor 414-1. In this process, the blade 511 and the blade 512 included in the first blocking unit 510 may be connected to the blade 411 and the blade 412 so as to move in synchronization, respectively. In some embodiments, the blade 421 and the blade 422 included in the second iris unit 420 may be moved by receiving driving power generated by the second driving motor 424-1. In this process, the blade 521 and the blade 522 included in the second blocking unit 520 may be connected to the blade 421 and the blade 422 so as to move in synchronization, respectively. Hereinafter, for convenience of explanation, movements of the blade 411 and the blade 412 and the and blades 511 and 521 will be described by mainly focusing on the blade 411 and the blade 412 included in the first iris unit 410 and the blade 511 and the blade 521 included in the first blocking unit 510. Referring to FIGS. 8 and 9, as driving power is generated by the first driving motor 414-1, an output axis 414-11 may rotate. The timing belt 415-11 which is supported by the output axis 414-11 and the support axis 415-12 so as to revolve may revolve in the same direction as the output axis 414-11 rotates. In various embodiments, in the case that the output axis 414-11 rotates clockwise, the timing belt 415-11 may also revolve clockwise. As the timing belt 415-11 revolves clockwise, a first connecting member 416-1 and a second connecting member 416-2 arranged to be fixed to the timing belt 415-11 may move in different directions along the X-axis. In this case, as the first connecting member 416-1 and the second connecting member 416-2 are arranged to be fixed to one timing belt 415-11, the two may move in synchronization. Accordingly, as the timing belt 415-11 revolves, the first connecting member 416-1 and the second connecting member 416-2 may move in different directions but a same distance. The blade 411 and the blade 412 fixed to each of the first connecting member 416-1 and the second connecting member 416-2 may move in different directions along the X-axis together with the first connecting member 416-1 and the second connecting member 416-2. As the blade 411 and the blade 412 move in different directions along the X-axis, a separation distance may be adjusted between the blade 411 and the blade 412, and the first field size range P made by the blade 411 and the blade 412 along the X-axis may also be adjusted. As the adjustment of the first field size range P along the Z-axis caused by the movement of the second driving motor 424-1 and the blade 421 and the blade 422 is virtually the same as the adjustment of the first field size range P along the X-axis caused by the movement of the blade 411 and the blade 412, explanations thereof will be omitted herein. As previously described in detail, as the blade 411 and the blade 412 move along the X-axis, the second slide unit 642 connected to be fixed to each of the blade 411 and the blade 412 may also move along the X-axis, by using the joint 643. In this process, the second slide unit 642 may slide along the direction of extension of the second long hole 641. In some embodiments, as the second slide unit 642 slides along the direction of extension of the second long hole 641, the connecting unit 620 may revolve around the hinge unit 610. As the connecting unit 620 revolves around the hinge unit 610, the first slider 632 inserted in the first long hole 631 may also slide along the direction of extension of the first long hole 631. As a result, the blade 511 and the blade 512 arranged to be fixed to the first slider 632 may slide along the X-axis, and it may result in an adjustment of the second field size range T. In various embodiments, as the blade 411 moves toward the blade 412 along the X-axis, the second slide unit 642 arranged to be fixed to the blade 411 may also move toward the blade 412 along the X-axis. In this case, the second slide unit 642 may slide to move closer to the hinge unit 610 along the direction of extension of the second long hole 641. As a result, the connecting unit 620 may revolve around the hinge unit 610 counter-clockwise. As the connecting unit 620 revolves around the hinge unit 610 counter-clockwise, the first slider 632 inserted in the first long hole 631 may slide to become closer to the hinge unit 610 along the direction of extension of the first long hole 631. As a result, the blade 511 arranged to be fixed to the first slider 632 may move to become close to the blade 512 along the X-axis, and as a result it may adjust the second field size range T. As the movement of the blade 521 and the blade 522 connected to the blade 421 and the blade 422 so as to work in synchronization with each other is virtually the same as that of the blades 421 and 422, explanations thereof will be omitted herein. As the movement of the blades 411, 412, 421 and 422 works in synchronization with the blades 511, 512, 521 and 522, the first field size range P and the second field size range T may be adjusted in proportion to each other. In various embodiments, as the blade 411 moves a first moving distance A along the X-axis, the second slide unit 642 arranged to be fixed to the blade 411 moves the first moving distance A along the X-axis. In this case, the first slider 632 arranged to work in synchronization with the second slide unit 642 due to the connecting unit 620 may move a second moving distance B along the X-axis. As a result, the blade 511 arranged to be fixed to the first slider 632 may also move the second moving distance B along the X-axis. In addition, in the path of moving of the connecting unit 620, the first slider 632 and the second slide unit 642 caused by the movement of the blade 411 and the blade 511, the first moving distance A and the second moving distance B of the first slider 632 and the second slide unit 642 may be proportionate to a first separation distance C and a second separation distance D from the hinge unit 610 to the first and the second sliders 632 and 642. For example, a ratio of the first separation distance C from the hinge unit 610 to the second slide unit 642 to the second separation distance D from the hinge unit 610 to the first slider 632 may be the same as the first moving distance A of the second slide unit 642 moving along the X-axis and the second moving distance B of the first slide unit 632 moving along the X-axis. Accordingly, in the case that the ratio of the first separation distance C and the second separation distance D from the hinge unit 610 to the first slider 632 and the second slider 642 is adjusted, a ratio of the first field size range P to the second field size range T which may be formed by the blade 411 and the blade 511 arranged to work in synchronization with each other may be adjusted. When moving the blade 411 and the blade 412 and the blade 511 and the blade 512 connected to the blade 411 and the blade 412 by using the timing belt 415-11 connected to the first driving motor 414-1, a symmetric adjustment alone is possible for the first field size range P and the second field size range T because the blade 411 and the blade 412 may move in synchronization with each other. By contrast, in the case that the driving motor or the timing belt to drive the blade 411 and the blade 412 is provided, it is easier to adjust the first field size range P and the second field size range T, because the blade 411 and the blade 412 may move more independently. FIG. 10 is a perspective view illustrating the collimator, according to another embodiment. FIG. 11 is a perspective view illustrating the first iris unit and the first blocking unit included in the collimator as shown in FIG. 10, according to an embodiment. For the convenience of explanation, descriptions of virtually the same structures as those described in FIGS. 5 and 6 will be omitted herein. Referring to FIGS. 10 and 11, a first iris unit 410 according to another embodiment may include a blade 411 that may move along the X-axis, a blade 412, a first slide support unit 413 that may support the blade 411 and the blade 412, a driving motor 414-1a generating driving power for moving each of the blade 411 and the blade 412, a driving motor 414-1b, and the 1st driving power transfer unit 415-1 and the driving power transfer unit 415-2 that may transfer driving power generated by the driving motor 414-1a and the driving motor 414-1b to the blade 411 and the blade 412. The driving motor 414-1a and the driving motor 414-1b are driving members that may move the blade 411 and the blade 412. The driving power transfer unit 415-1 may be connected to a first output axis 414-11a of the driving motor 414-1a while the driving power transfer unit 415-2 may be connected to a second output axis 414-11b of the driving motor 414-1b. The driving power transfer unit 415-1 and the driving power transfer unit 415-2 are driving power transferring members that may transfer driving power generated by the driving motor 414-1a and the driving motor 414-1b to the blade 411 and the blade 412, respectively. In various embodiments, the driving power transfer unit 415-1 may include the first timing belt 415-11a which may move due to a revolution of the first output axis 414-11a while being arranged to face the first output axis 414-11a and the first support axis 415-12a that may support the first timing belt 415-11a while being arranged to face the first output axis 414-11a. In some embodiments, the driving power transfer unit 415-2 may include the second timing belt 415-11b which may move due to a rotation of the second output axis 414-11b while being arranged to face the second output axis 414-11b and the second support axis 415-12b that may support the second timing belt 415-11b while being arranged to face the second output axis 414-11b. The first timing belt 415-11a may be arranged to revolve around the first output axis 414-11a and the first support axis 415-12a while being supported by the first output axis 414-11a and the first support axis 415-12a. In some embodiments, the second timing belt 415-11b may be arranged to revolve around the second output axis 414-11b and the second support axis 415-12b while being supported by the second output axis 414-11b and the second support axis 415-12b. In this case, the first connecting member 416-1 and the second connecting member 416-2 may be arranged to be fixed to the first timing belt 415-11a and the second timing belt 415-11b, respectively. As a result, the first connecting member 416-1 and the second connecting member 416-2 may be moved individually along the path of the first timing belt 415-11a and the second timing belt 415-11b. The blade 411 and the blade 412 which are arranged to be fixed to the first connecting member 416-1 and the second connecting member 416-2 may also be moved along the path of the first timing belt 415-11a and the second timing belt 415-11b. The second iris unit 420 may include the blade 421 that may move along the Z-axis, the blade 422, the second slide support unit 423 that may support the blade 421 and the blade 422, a driving motor 424-1a generating driving power for moving each of the blade 421 and the blade 422, a driving motor 424-1b, and the 2-1st driving power transfer unit 425-1 and the driving power transfer unit 425-2 that may transfer driving power generated by the 2-1st driving motor 424-1a and the driving motor 424-1b to the blade 421 and the blade 422. As the 2-1st driving motor 424-1a, the driving motor 424-1b, the 2-1st driving power transfer unit 425-1 and the driving power transfer unit 425-2 which are included in the second iris unit 420 may be virtually the same as the 1-1st driving motor 414-1a, the driving motor 414-1b, the 1-1st driving power transfer unit 415-1 and the driving power transfer unit 415-2 of the first iris unit 410, explanations thereof will be omitted herein. FIG. 12 is a block diagram illustrating a structure of the collimator, according to another embodiment. FIG. 13 is a front view illustrating the first iris unit which shows a movement status of the blades 1-1 and 2-2, according to another embodiment. Referring to FIG. 12, the 1-1st through blades 411, 412, 421 and 422 included in the first iris unit 410 and the second iris unit 420 may be independent from each other as they receive driving power generated by the 1-1st driving motor 414-1a, the 1-2nd driving motor 414-1b, the 1-3rd driving motor 424-1a, and the 1-4th driving motor 424-1b. In this case, the blades 511, 512, 521 and 522 included in a first blocking unit 510 and a second blocking unit 520 may be connected to move in synchronization with the 1-1st through blades 411, 412, 421 and 422. Hereinafter, for convenience of explanation, movements of the blade 411 and the blade 412 and the blade 511 and the blade 521 will be described, by mainly focusing on the blade 411 and the blade 412 included in the first iris unit 410 and the 2-1st blade 511 and the blade 521 included in the first blocking unit 510. Referring to FIG. 13, as driving power is generated by the 1-1st driving motor 414-1a and the 1-2nd driving motor 414-1b, the first output axis 414-11a and the second output axis 414-11b may revolve. The first timing belt 415-11a supported to revolve by the first output axis 414-11a and the first support axis 415-12a, and the second timing belt 415-11b supported to revolve by the second output axis 414-11b and the second support axis 415-12b may revolve in the same direction as the first output axis 414-11a and the second output axis 414-11b with the first output axis 414-11a and the second output axis 414-11b rotating. In various embodiments, in the case that the first output axis 414-11a rotates clockwise, the first timing belt 415-11a may also revolve clockwise, and in the case that the second output axis 414-11b rotates counter-clockwise, the second timing belt 415-11b may also revolve counter-clockwise. As the first timing belt 415-11a revolves clockwise and the second timing belt 415-11b revolves counter-clockwise, the first connecting member 416-1 and the second connecting member 416-2, arranged to be fixed to the first timing belt 415-11a and the second timing belt 415-1b respectively, may move in different directions along the X-axis. In this case, the first connecting member 416-1 and the second connecting member 416-2 may move independently from each other because the first connecting member 416-1 and the second connecting member 416-2 are fixed to the first timing belt 415-11a and the second timing belt 415-1b, respectively. That is, as the first timing belt 415-11a and the second timing belt 415-11b revolve in different directions, the first connecting member 416-1 and the second connecting member 416-2 may move different distances in different directions along the axis X. In some embodiments, in the case that the first timing belt 415-11a and the second timing belt 415-11b revolve in the same direction, it is obvious that the first connecting member 416-1 and the second connecting member 416-2 may move in the same direction along the axis X. The blade 411 and the blade 412 arranged to be fixed to the first connecting member 416-1 and the second connecting member 416-2 may move independently from each other along the X-axis together with the first connecting member 416-1 and the second connecting member 416-2. As the 1st blade 411 and the blade 412 move independently from each other along the X-axis, the 1st blade 511 and the blade 512 connected to move in synchronization with the 1st blade 411 and the blade 412 may also move independently from each other. As a result, the separation distance between the 1-1st blade 411 and the blade 412 and the blade 511 and the blade 512 may be adjusted more precisely, and the first field size range P caused by the 1st blade 411 and the blade 412 along the X-axis and the second field size range T caused by the blade 511 and the blade 512 along the X-axis may be adjusted more precisely. FIG. 14 is a flowchart illustrating a method of regulating the irradiation range of the X-ray imaging apparatus. Referring to FIGS. 1, 6, 7 and 14, in operation 210, an adjustment signal on the first field size range P and the second field size range T is received. (S210) In the X-ray imaging apparatus according to an embodiment, a user may receive the adjustment signal with respect to the first field size range P and the second field size range T formed by the collimator 300 through an input unit 142. In operation S220, upon receiving the adjustment signal, a driving signal for the first driving motor 414-1 and the second driving motor 424-1 may be generated. (S220) A controller 15 may generate the driving signal for the first driving motor 414-1 and the second driving motor 424-1 by using the adjustment signal received through the input unit 142 with respect to the first field size range P and the second field size range T. In operation S230, the first driving motor 414-1 and the second driving motor 424-1 may generate driving power, upon receiving the driving signal. (S230) The first driving motor 414-1 and the second driving motor 424-1 may generate the driving power, upon receiving the driving signal transferred from the controller 15. In this case, driving signals to drive the first driving motor 414-1 and the second driving motor 424-1 may be different from each other. As a result, driving powers generated by the first driving motor 414-1 and the second driving motor 424-1 may also be different from each other. In operation S240, the driving power generated by the first driving motor 414-1 and the second driving motor 424-1 may be transferred to the blade 411 and the blade 412 and the blade 421 and the blade 422, respectively. (S240) The driving power generated by the first driving motor 414-1 and the second driving motor 424-1 may be transferred to the blade 411 and the blade 412 and the blade 421 and the blade 422 through the first driving power transfer unit 415 and the second driving power transfer unit 425. In this case, the 1-1st blade 411 and the blade 412 may receive the driving power generated by the first driving motor 414-1 through the first driving power transfer unit 415, and the blade 421 and the blade 422 may receive the driving power generated by the second driving motor 424-1 through the second driving power transfer unit 425. In operation S250, the blade 411 and the blade 412 and the blade 421 and the blade 422 may move. (S250) The blade 411 and the blade 412 which received the driving power generated by the first driving motor 414-1 through the first driving power transfer unit 415 may move in synchronization with each other, and the blade 421 and the blade 422 which received the driving power generated by the second driving motor 424-1 through the second driving power transfer unit 425 may also move in synchronization with each other. However, as previously described in detail, as the driving power generated at the first driving motor 414-1 and the second driving motor 424-1 may be different from each other, the blade 411 and the blade 412 and the blade 421 and the blade 422 may move independently. In operation 260, the blade 511 and the blade 512 may move in synchronization with the blade 411 and the blade 412, and the blade 521 and the blade 522 may also move in synchronization with the blade 421 and the blade 422. (S260) As the blade 411 and the blade 412 move, the blade 511 and the blade 512 arranged to move in synchronization with the blade 411 and the blade 412 may be moved, by using the connector 600. In some embodiments, as the blade 421 and the blade 422 move, the blade 521 and the blade 522 arranged to move in synchronization with the blade 421 and the blade 422 may be moved, by using the connector 600. As a result, by driving just the first driving motor 414-1 and the second driving motor 424-1, which may drive the blade 411 and the blade 412 and the blade 421 and the blade 422, it is possible to adjust the movement range of the blade 511 and the blade 512 and the blade 521 and the blade 522, and thus, it is possible to adjust the first field size range P and the second field size range T. FIG. 15 is a flowchart illustrating a method of regulating the irradiation range for the X-ray imaging apparatus, according to another embodiment. Referring to FIGS. 1, 10, 11 and 13, in operation S310, an adjustment signal with respect to the first field size range P and the second field size range T is received. (S310) In the X-ray imaging apparatus according to an embodiment, a user may receive an adjustment signal with respect to the first field size range P and the second field size range T formed by the collimator 300 through an input unit 142. In operation S320, a driving signal for a plurality of driving motors is generated depending on the received adjustment signal. (S320) The controller 15 may generate a driving signal for the plurality of driving motors by using an adjustment signal received with respect to the first field size range and the second field size range through the input unit 142. The plurality of driving motors are the 1-1st driving motor 414-1a, the driving motor 414-1b, the 2-1st driving motor 424-1a and the driving motor 424-1b which may drive the blades 411, 412, 421 and 422, respectively. In operation S330, the plurality of driving motors may generate driving power, upon receiving a driving signal. (S330) The plurality of driving motors may generate driving power, upon receiving the driving signal transferred from the controller 15. In this case, the driving signal to drive the plurality of driving motors (i.e., the 1-1st driving motor 414-1a, the driving motor 414-1b, the 2-1st driving motor 424-1a, and the driving motor 424-1b) may be different from each other. As a result, driving power generated by the 1-1st driving motor 414-1a, the driving motor 414-1b, the 2-1st driving motor 424-1a, and the driving motor 424-1b may also be different from each other. In operation S340, driving power generated by the plurality of driving motors may be transferred to the blades 411, 412, 421 and 422, respectively. (S340) Driving power generated by the plurality of driving motors (i.e., the 1-1st driving motor 414-1a, the driving motor 414-1b, the 2-1st driving motor 424-1a, and the driving motor 424-1b) may be transferred to the blades 411, 412, 421 and 422. In this case, the blades 411, 412, 421, and 422 may be transferred driving power generated by the plurality of driving motors through the plurality of the driving power transfer units. In operation S350, the blades 411, 412, 421 and 422 may move. (S350) Driving power generated by the plurality of driving motors (i.e., the 1-1st driving motor 414-1a, the driving motor 414-1b, the 2-1st driving motor 424-1a, and the driving motor 424-1b) may be different from each other. As a result the blades 411, 412, 421 and 422 may move independently from each other. In operation S360, the blades 511, 512, 521 and 522 may move in synchronization with the blades 411, 412, 421 and 422. (S360) As the blades 411, 412, 421 and 422 move, the 1st blade 511 and the blade 512 arranged to move in synchronization with the blades 411, 412, 421 and 422 may be moved, by using the connector 600. As a result, by driving just the plurality of driving motors (i.e., the 1-1st driving motor 414-1a, the driving motor 414-1b, the 2-1st driving motor 424-1a, and the driving motor 424-1b), which may drive the blades 411, 412, 421 and 422, it is possible to adjust the movement range of the blades 511, 512, 521 and 522, and thus, it is possible to adjust the first field size range P and the second field size range T. The X-ray imaging apparatus according to an embodiment may prevent image defects, which may be caused by such problems as a control error, by using a plurality of X-ray field size range adjustment units which are arranged to move in synchronization with each other. In addition, it is possible to improve manufacturing convenience and reduce manufacturing costs, as the plurality of X-ray field size range adjustment units may be adjusted by using one driving motor. Although the X-ray imaging apparatus and the method of operating the X-ray imaging apparatus according to one or more embodiments have been described with reference to the figures, the aforementioned embodiments are merely examples. 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 present invention as defined by the following claims.
	claims	1. An optical device for directing x-rays, the optical device comprising:a plurality of curved optical crystals, each with at least one lattice plane, positioned according to an x-ray source and an x-ray target to define at least one Rowland circle of radius R and a source-to-target line;wherein the optical device provides focusing of x-rays from the source to the target;wherein the plurality of curved optical crystals have a radius at the plane of the Rowland circle different than R; andwherein an angle of at least one lattice plane of a first crystal of the plurality of curved optical crystals relative to a surface of the first crystal is different from an angle of at least one lattice plane of a second crystal of the plurality of curved optical crystals relative to surface of the least one second crystal. 2. The optical device of claim 1, wherein the plurality of curved optical crystals have a radius at the plane of the Rowland circle of about 2R. 3. The optical device of claim 2, wherein at least one of the plurality of optical crystals comprises a surface upon which x-rays are directed, and wherein at least one of the plurality of optical crystals comprises a set of atomic diffraction planes having a Bragg angle θB and oriented at an angle γ with the surface of the at least one of the plurality of optical crystals, and wherein a line drawn from the x-ray source to a point on the surface of the at least one of the plurality of optical crystals makes an angle θB+γ with the source-to-target line. 4. The optical device of claim 3, wherein the line drawn from the x-ray source to a point on the surface of the at least one of the plurality of optical crystals comprises a line drawn from the x-ray source to the midpoint of the surface of the at least one of the plurality of optical crystals. 5. The optical device of claim 3, wherein the line drawn from the x-ray source to a point on the surface of the at least one of the plurality of optical crystals comprises a line drawn from the x-ray source to about the point of tangency of the surface of the at least one of the plurality of optical crystals and the Rowland circle of the at least one of the plurality of optical crystals. 6. The optical device of claim 1, wherein at least one of the crystals is a doubly-curved crystal. 7. The optical device of claim 6, wherein at least one of the crystals is a toroidal doubly-curved crystal. 8. The optical device of claim 1, in combination with a source of x-rays. 9. The optical device of claim 1, wherein the optical device comprises a toroidal angle about the source-to-target line of at least about 90 degrees. 10. The optical device of claim 9, wherein the optical device comprises a toroidal angle about the source-to-target line of at least about 180 degrees. 11. The optical device of claim 10, wherein the optical device comprises a toroidal angle about the source-to-target line of at least about 270 degrees. 12. The optical device of claim 11, wherein the optical device comprises a toroidal angle about the source-to-target line of about 360 degrees. 13. The optical device of claim 1, wherein the angle of the lattice planes of the first crystal relative to the surface of the first crystal is about zero. 14. The optical device of claim 1, wherein the angle of the at least one lattice plane of the at least one second crystal relative to the surface of the at least one second crystal is at least about 5 degrees greater than the angle of the at least one lattice plane of the first crystal relative to the surface of the first crystal. 15. The optical device of claim 1, wherein a line directed from the x-ray source to the center of a surface of the first curved crystal and a line directed from the x-ray source to the center of a surface of the at least one second crystal define an angle γ. 16. The optical device of claim 15, wherein the angle of the at least one lattice plane of the at least one second crystal relative to the surface of the at least one second crystal is about γ. 17. The optical device of claim 1, wherein the angle γ is between about minus 15 degrees and about plus 15 degrees. 18. A method for directing x-rays, comprising:providing an optical device, the optical device comprising a plurality of curved optical crystals, each with at least one lattice plane, arranged in a matrix, the matrix being positionable to define at least one Rowland circle of radius R with an x-ray source and an x-ray target, and wherein the matrix comprises a plurality of rows, with each row comprising multiple optical crystals of said plurality of optical crystals, wherein the plurality of curved optical crystals have a radius at the plane of the Rowland circle different than R; andpositioning the optical device wherein at least some x-rays from the x-ray source are directed to the x-ray target;wherein an angle of at least one lattice plane of a first crystal of the plurality of curved optical crystals relative to a surface of the first crystal is different from an angle of at least one lattice plane of a second crystal of the plurality of curved optical crystals relative to a surface of the at least one second crystal. 19. The method of claim 18, wherein the plurality of curved optical crystals have a radius at the plane of the Rowland circle of about 2R. 20. The method of claim 18, wherein positioning the optical device comprises positioning the device wherein at least some x-rays emitted by the source impinge at least some of the crystals of the optical device wherein at least some of the x-rays are diffracted. 21. A device for directing x-rays, comprising a first curved crystal and at least one second curved crystal spaced from the first crystal, the first and at least one second curved crystal each comprising at least one lattice plane, and the first curved crystal and the at least one second curved crystal being positionable to define at least one Rowland circle of radius R with an x-ray source and an x-ray target, wherein at least some x-rays impinging upon the first curved crystal and the at least one second curved crystal are directed to the target, and wherein an angle of the at least one lattice plane of the first crystal relative to a surface of the first crystal is different from an angle of the at least one lattice plane of the at least one second crystal relative to a surface of the at least one second crystal. 22. The device of claim 21, wherein the angle of the lattice planes of the first crystal relative to the surface of the first crystal is about zero. 23. The device of claim 21, wherein the angle of the at least one lattice plane of the at least one second crystal relative to the surface of the at least one second crystal is at least about 5 degrees greater than the angle of the at least one lattice plane of the first crystal relative to the surface of the first crystal. 24. The device of claim 21, wherein a line directed from the x-ray source to the center of a surface of the first curved crystal and a line directed from the x-ray source to the center of a surface of the at least one second crystal define an angle γ. 25. The device of claim 24, wherein the angle of the at least one lattice plane of the at least one second crystal relative to the surface of the at least one second crystal is about γ. 26. The device of claim 25, wherein the angle γ is between about minus 15 degrees and about plus 15 degrees. 27. The device of claim 21, wherein the first curved crystal and the at least one second crystal comprise a first set of crystals, and the device further comprises at least one second set of crystals which are also positioned to define a Rowland circle with the x-ray source and the x-ray target, wherein at least some x-rays which impinge upon the at least one second set of crystals are directed to the x-ray target, the target being common with the first set of crystals, and wherein the second set of crystals is spaced from the first set of crystals in a direction orthogonal to a plane of the Rowland circle of the first set of crystals.
	claims	1. An intermediate end plug assembly for a segmented fuel rod comprising:a first end plug having an outer cylindrical body including an inner cylindrical body, in a center of which a through-hole is formed in a longitudinal direction, the outer cylindrical body having an annular groove in one of circular upper and lower surfaces thereof between an outer circumference thereof and an outer circumference of the inner cylindrical body, and an annular protrusion protruding along an outer circumference thereof and a first flat coupling face inside the protrusion on the other surface thereof; anda second end plug having a cylindrical member, which has a body and a through-hole having diameters identical to those of the respective body and through-hole of the first end plug, the body of the second end plug having a groove identical to the groove of the first end plug in one of circular upper and lower surfaces thereof, and a cylindrical insert having an annular space along an outer circumference thereof so as to correspond to the annular protrusion of the first end plug and a second flat coupling face on an upper surface of the insert on a second surface thereof,wherein the protrusion of the first end plug is inserted into the annular space of the second end plug, so that the first and second coupling faces come into close contact with each other,wherein the annular protrusion has “L” shaped coupling recesses, each of which is partially open, and the insert of the second end plug has latches fitting into the respective coupling recesses on an outer circumference thereof,wherein the latches protrude in the outside direction of the insert,wherein each coupling recess includes a seat having a locking step protruding in a predetermined height,wherein, in comparison of inner and outer annular faces located inside and outside grooves of the first and second end plugs, the inner annular face protrudes higher than the outside annular face. 2. The intermediate end plug assembly as set forth in claim 1, wherein the coupling recesses and the latches are equal to each other in number, and the latches are formed on the outer circumference of the inset of the second end plug so as to correspond to a position where the coupling recesses are formed. 3. The intermediate end plug assembly as set forth in claim 1, wherein, in comparison of inner and outer annular faces located inside and outside grooves of the first and second end plugs, the former protrudes higher than the latter. 4. The intermediate end plug assembly as set forth in claim 1, wherein at least one of the first and second end plugs has at least one complementary channel hole communicating with the through-hole. 5. A dual-cooled fuel rod comprising:a segmented upper fuel rod;an upper intermediate end plug connected to the segmented upper fuel rod;a lower intermediate end plug connected to the upper intermediate end plug; anda segmented lower fuel rod connected to the lower intermediate end plug,wherein the upper intermediate end plug has a cylindrical member, which has a body and a through-hole having diameters identical to those of the respective body and through-hole of the lower intermediate end plug, the body of the upper intermediate end plug having a groove identical to the groove of the lower intermediate end plug in one of circular upper and lower surfaces thereof, and a cylindrical insert having an annular space along an outer circumference thereof so as to correspond to the annular protrusion of the lower intermediate end plug and a second flat coupling face on an upper surface of the insert on a second surface thereof,wherein the lower intermediate end plug has an outer cylindrical body including an insert cylindrical body, in a center of which a through-hole is formed in a longitudinal direction, the outer cylindrical body having an annular groove in one of circular upper and lower surfaces thereof between an outer circumference thereof and an outer circumference of the inner cylindrical body, and an annular protrusion protruding along an outer circumference thereof and a first flat coupling face inside the protrusion on the other surface thereof,wherein the protrusion of the upper intermediate end plug is inserted into the annular space of the lower intermediate end plug, so that the first and second coupling faces come into close contact with each other,wherein the annular protrusion has “L” shaped coupling recesses, each of which is partially open, and the insert of the lower intermediate end plug has latches fitting into the respective coupling recesses on an outer circumference thereof,wherein the latches protrude in the outside direction of the insert,wherein each coupling recess includes a seat having a locking step protruding in a predetermined height,wherein, in comparison of inner and outer annular protrusions located inside and outside grooves of the upper intermediate and lower intermediate end plugs, the inner annular protrusion protrudes higher than the outside annular protrusion. 6. The dual-cooled fuel rod as set forth in claim 5, wherein at least one of the upper intermediate end plug and lower intermediate end plug has at least one complementary channel hole, which communicates with the through-hole and is inclined toward the segmented upper fuel rod. 7. The dual-cooled fuel rod as set forth in claim 5, wherein each of the segmented upper and lower fuel rods has a plenum spring and a spacer installed in an inner annular space thereof. 8. The dual-cooled fuel rod as set forth in claim 5, wherein each of the segmented upper and lower fuel rods has an elongation ratio ranging from 100 to 200.
059986905	description	EXAMPLE 1 288 parts by weight of 95% sodium hydroxide and 1,400 parts by weight of 99% boric acid were obtained and were each divided into two equal parts. Each of the two equal parts was again divided twice and the parts were added in order gradually into 600 parts by weight of deionized water under agitation. The sequence of addition is as follows: sodium hydroxide--boric acid--sodium hydroxide--boric acid. To wait for sodium hydroxide was completely dissolved, the mixture solution was heated slightly to allow boric acid to dissolve completely. Dissolved boron concentration of the solution thus obtained was 105,943 ppm and the molar ratio of sodium/boron was 0.3. After boric acid was dissolved, the solution was continuously stirred and was cooled to 40.degree. C., at which temperature the solution was kept for ready use. Before addition of the solidification agent, the solution must be weighed again in order to know the weight lost by evaporation of moisture in the abovementioned preparation process and was supplemented with water of the same temparature. 16 parts of Portland type II cement produced by Taiwan Cement Company, 13 parts of tribasic magnesium phosphate powder and 0.4 part of stranded carbon fiber were mixed, homogenized and then pulverized to prepare the solidification agent powder. Thereafter, this solidification agent powder was gradually added into a ready-for-use boric acid solution and was at the same time vigorously agitated to allow the solidification agent powder to mix with the solution to form a homogeneous slurry. The weight ratio of solidification agent to waste fluid is 0.4. Agitation was stopped ten minutes after the solidification agent was completely added, the slurry was immediately poured into a cylindrical polyethylene plastic model having an inner diameter of 5 cm and a height of 11 cm and then was left at the room temperature. Demolding took place 30 days after the solidification and 5 samples were obtained and cut into 10 cm long cylindrical specimens, the specimens were again tested for compressive strength under ASTM C39 procedure in accordance with the quality specification of the U.S. Nuclear Regulatory Commission. From the result of the test, the average compressive strength of the 5 samples is 189 kg/cm.sup.2. EXAMPLE 2 Borate solution and solidification agent were prepared in the same steps as in Example 1. In the solution, the concentration of the dissolved boron and the molar ratio of sodium: boron were also the same as in Example 1; the component of solidification agent was however changed to 4 parts of type 2A mud solidification agent (for composition, please refer to U.S. Pat. No. 5,457,262) with 1 part of magnesium oxide, 1 part of tribasic magnesium phosphate and 0.09 part of stranded carbon fiber. The weight ratio of solidification to liquid waste used was 0.3328. Demolding took place 7 days after the solidification and test was performed similarly with 5 samples. From the result, the compressive strength is 130 kg/cm.sup.2. EXAMPLE 3 Borate solution and solidification agent were prepared in the same steps as in Example 1. In the solution, the concentration of the dissolved boron and the molar ratio of sodium:boron were the same as in Example 1; the component of solidification agent was however changed to 15 parts of Portland cement with 3 parts of fume silica, 7 parts of silicon phosphate and 0.4 part of carbon fiber. The weight ratio of solidification agent to liquid waste used in the solidification was lowered to 0.289. From the result, it was obtained that the compressive strength after presservation for 8 months of the solidified body is 105 kg/cm.sup.2 and the water resistant compressive strength is 93 kg/cm.sup.2. EXAMPLE 4 Borate solution was prepared in the same steps as in Example 1 and in the solution the concentration of boron was made to be 120,000 ppm and the molar ratio of sodium:boron was 0.32. Thereafter, the fine powder of BaSiO.sub.3 was used as the solidification agent and solidification was performed with a ratio of each part of borate solution with 0.37 part of solidification agent. Demolding took place 7 days after solidification and test was performed similarly with 5 samples. From the result, the compressive strength is 61 kg/cm.sup.2. EXAMPLE 5 Borate solution was prepared in the same steps as in Example 1, however, the molar ratio of sodium:boron was raised and pH of the solution was adjusted low with 85% phosphoric acid. The prepared simulative liquid borate waste was measured to contain boron of 77,728 ppm, the molar ratio of sodium:boron of 0.7 and phosphoric acid (H.sub.3 PO.sub.4) of 25,909 ppm. The preparation process of the solidification agent was also the same as in Example 1, its composition being 13 parts of type IIA mud solidification agent of Taiwan cement Company with 6 parts of magnesium oxide and 0.3 part of stranded carbon fiber. In solidification, the weight ratio of solidification agent to liquid waste was 0.2383. Demolding took place 30 days after solidification and test was performed similarly with 5 samples. From the result, the compressive strength is 193 kg/cm.sup.2 and the water resistant compressive strength is 172 kg/cm.sup.2.
046860681	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention concerns a method of treating radioactive wastes mainly composed of various combustible and/or poorly combustible organic materials, or mixtures thereof with inorganic materials. 2. Description of the Prior Art Various kinds of radioactive organic waste are generated in nuclear power plants and facilities handling radioactive substances. At present, some of these wastes are partially treated, but most are left untreated. That is, the waste is charged and stored in an untreated state in drums or other containers. Filtration aids, spent ion exchange resins and the like, are stored, as they are, along with liquid waste in storage tanks. The amount of radioactive wastes generated will increase over time, and the places or facilities for storing accumulated radioactive wastes will inevitably be filled up. Furthermore, when solid radioactive wastes are stored, as they are, mixed with liquid waste in storage tanks, safety problems may arise. Accordingly, there is an urgent demand for developing end establishing a practical treating method for solving these problems. Various methods of treating radioactive wastes, such as, compaction, incineration and acid digestion, are now under development or have been put to practical use. However, these waste treatment methods have serious drawbacks which limit their utility. The compaction method does not exhibit an effective volume-reduction ratio of the treated radioactive wastes, and therefore does not satisfactorily minimize the required waste storage space. Furthermore, the substances which can be processed by the compaction method are limited. Referring to the incineration method, there is an additional requirement for facilities to seperate and remove sulfur oxides, nitrogen oxides, chlorine, hydrogen chloride and fine ash that are contained in a large volume in the gaseous wastes generated during incineration. Also, secondary radioactive wastes are generated from the treatment or the geseous wastes. Therefore, in the incineration method, the volume-reduction effect of the treatment is offset by the above named problems and another problem is caused by corrosion or the incineration facilities. There has been well known a wet oxidation process which is employed in the treatment of liquid wastes containing ordinary organic substances, in which the liquid wastes are continuously introduced into an oxidizing reactor at a high temperature end high pressure end are subjected to air oxidation. However, the oxidizing reaction tends to be terminated at the formation of organic acids in this method. As a means for avoiding this defect and improving the degree of oxidation, it has also been known that the addition of a catalyst, such as copper ions, is effective for oxidizing the organic substances more completely. It may be possible to apply such a conventional continuous wet oxidizing process to radioactive organic wastes. However, there have been various drawbacks such as (1) the degree or oxidation or the organic meterials is low, (2) a great amount of waste water containing radioactive substances end heavy metals used as the catalyst are discharged thereby creating various problems in the treatment or the discharged waste water and (3) if thermoplastic high molecular polymeric substances are contained in the radioactive wastes, they will be melted in the reactor and fused together (hereinefter referred to as fusion) with other organic wastes, thereby forming large lumps which hinder the progress of the oxidizing treatment. Because of these drawbacks, application of the wet oxidizing process to treat radioactive wastes has not yet been put to practical use. It is usually difficult to treat a waste mixture, at the location at which it is generated, to separate it into its respective kinds of components and discharge them separately. Similarly, it is more difficult, if not impossible, to subsequently treat a waste mixture to separate those different kinds or components. If the wastes contain naturally radioactive substances, this difficulty is increased more remerkably. SUMMARY OF THE INVENTION Accordingly, it is a first object of this invention to provide a method of treating a mixture of organic radioactive wastes, which mixture may or may not also contain inorganic wastes. The second object or this invention is to provide a method of treating radioactive wastes containing thermoplastic, high molecular weight, polymeric materials without seperating or removing such polymeric materials in advance, even though such polymeric materials have heretofore caused problems in carrying out the treatment or radioactive wastes by the wet oxidizing process. The third object or this invention is to provide a method of treating radioactive wastes by wet oxidization in which the only required pre-treatment is to cut the waste materials into a predetermined size and configuration so that they can be fed into an oxidizing reactor. The fourth object or this invention is to increase the volume-reduction ratio achieved by the treatment of the radioactive organic wastes and, as a result, minimize the volume or the final waste to be stored. The fifth object of this invention is to provide a method of oxidizing radioactive wastes safely and effectively and which requires a low amount of energy. These and other objects of the invention will become more apparent from a reading of the detailed description and examples which follow.
	claims	1. A light water reactor, comprising:a reactor pressure vessel that defines a volume;a core positioned in a bottom portion of the reactor pressure vessel and configured to support a plurality of nuclear fuel assemblies;a riser within the volume of the reactor pressure vessel, the riser extending from a position above the core and toward a top portion of the reactor pressure vessel; anda condensing steam generator positioned adjacent the riser and within the reactor pressure vessel volume, where a primary coolant comprises a liquid pooled at the bottom portion of the reactor pressure vessel, the pool having an upper surface positioned between the bottom portion of the reactor pressure vessel and a bottom of the condensing steam generator such that the condensing steam generator is not in contact with the liquid pooled at the bottom portion of the reactor pressure vessel during normal operation of the light water reactor and an upper portion of the volume located above the condensing steam generator includes a saturated steam dome absent a pressurizer,wherein the light water reactor is a self-pressurizing pressurized water reactor (PWR). 2. The light water reactor of claim 1, further comprising a primary coolant circuit entirely within the reactor pressure vessel volume, the primary cooling circuit extending along a path through the core, through an interior of the riser, along an exterior of the riser, and back through the core. 3. The light water reactor of claim 2, wherein a volume of the primary coolant and the primary coolant circuit are configured to guide a flow of the primary coolant through the primary coolant circuit at a saturation pressure of the primary coolant. 4. The light water reactor of claim 1, wherein the primary coolant is boron-free. 5. The light water reactor of claim 1, wherein the upper surface of the pool is positioned below the condensing steam generator and above the core. 6. The light water reactor of claim 5, further comprising:a containment vessel enclosing the reactor pressure vessel;a feed water input circuit extending through the containment vessel and the reactor pressure vessel to the condensing steam generator; anda steam output circuit extending through the containment vessel and the reactor pressure vessel to the condensing steam generator. 7. The light water reactor of claim 5, wherein the reactor pressure vessel is configured to contain pressures within a range between 1150 psia to 1750 psia. 8. The light water reactor of claim 3, further comprising a secondary coolant circuit thermally coupled to the primary coolant circuit through the condensing steam generator, the secondary coolant circuit controllable to maintain the flow of the primary coolant through the primary coolant circuit at the saturation pressure of the primary coolant. 9. The light water reactor of claim 1, wherein the reactor pressure vessel is pressurizer-less. 10. A light water reactor, comprising:a reactor pressure vessel that defines a volume, the reactor pressure vessel having a bottom portion and a top portion, the top portion forming an interior dome;a core positioned in the reactor pressure vessel and configured to support a plurality of nuclear fuel assemblies;a condensing steam generator positioned within the reactor pressure vessel volume between the core and the top portion of the reactor pressure vessel, the condensing steam generator defining an interior pathway through the condensing steam generator and an exterior annulus between the steam generator and a sidewall forming the reactor pressure vessel, where a liquid coolant surface level of a primary coolant is positioned below a bottom of the condensing steam generator and above the core such that the condensing steam generator is not in contact with the liquid coolant surface of the primary coolant during normal operation of the light water reactor and an upper portion of the volume located above the condensing steam generator includes a saturated steam dome absent a pressurizer;a primary coolant circuit that extends through the core, continues in a direction from the bottom portion toward the top portion through the interior pathway of the condensing steam generator, continues in a direction from the top portion toward the bottom portion through the exterior annulus, and returns to the core to recirculate through the primary coolant circuit; andthe primary coolant comprising a boron-free liquid,wherein the light water reactor is a self-pressurizing pressurized water reactor (PWR). 11. The light water reactor of claim 10, further comprising a riser extending above the core and toward the condensing steam generator. 12. The light water reactor of claim 10, further comprising:a containment vessel enclosing the reactor pressure vessel;a feed water input circuit extending through the containment vessel and the reactor pressure vessel to the condensing steam generator; anda steam output circuit extending through the containment vessel and the reactor pressure vessel to the condensing steam generator. 13. The light water reactor of claim 10, wherein the primary coolant comprises water. 14. The light water reactor of claim 10, wherein the primary coolant circuit is configured to enclose a primary coolant flow at a saturation pressure of the primary coolant. 15. The light water reactor of claim 14, further comprising a secondary coolant circuit thermally coupled to the primary coolant circuit through the condensing steam generator, the secondary coolant circuit controllable to maintain the flow of the primary coolant through the primary coolant circuit at the saturation pressure of the primary coolant. 16. A method for operating a light water reactor that is a self-pressurizing pressurized water reactor (PWR), the method comprising:flowing a primary coolant liquid at saturation pressure through a core that comprises a plurality of nuclear fuel assemblies;boiling the primary coolant liquid with heat from the plurality of nuclear fuel assemblies to form a primary coolant vapor;circulating the primary coolant vapor from above the core through a riser;condensing the primary coolant vapor on a steam generator positioned adjacent the riser to form the primary coolant liquid;circulating the primary coolant liquid, through an annulus defined between the riser and a reactor pressure vessel, to the core; andmaintaining a top water level of the primary coolant liquid at a level between a top of the core and a bottom of the steam generator such that the steam generator is not in contact with the top water level of the primary coolant liquid during normal operation of the self-pressurizing PWR and an upper portion of the reactor pressure vessel located above the steam generator operates absent a pressurizer. 17. The method of claim 16, wherein the primary coolant liquid comprises aboron-free liquid. 18. The method of claim 16, wherein the boiling occurs at a position below the steam generator. 19. The method of claim 16, further comprising transferring heat from the primary coolant vapor to a working fluid in the steam generator through a phase change in the primary coolant vapor to the primary coolant liquid. 20. The method of claim 19, further comprising:circulating the working fluid in the steam generator to a secondary coolant circuit that is fluidly coupled to a power generation system; andcontrolling the circulation of the working fluid in the secondary coolant circuit to maintain the flow of the primary coolant liquid at saturation pressure.
046506330	claims	1. In a nuclear power plant having a fluid-filled reactor vessel with a vapor outflow line for removing vapor from said reactor vessel, liquid inflow means for injecting liquid to said reactor vessel, said inflow means including an inflow line, a centrifugal pump disposed along said inflow line having an inlet and an outlet, an induction motor to drive said pump, flow control means along said inflow line between said pump and said reactor vessel from said pump, and means for generating a first control signal in response to liquid level in said reactor vessel and net vapor outflow versus liquid inflow with respect to said reactor vessel, said first control signal generating means being effective to generate a first signal to open and a second signal to close said flow control means to maintain liquid level in said vessel within predetermined limits, a pump and pump motor protection apparatus comprising: means for measuring the pressure of said liquid in the inlet of said pump; means for measuring the temperature of said liquid in the inlet of said pump; means for determining a required subcooling for said pump at the instantaneous temperature of said liquid in the inlet of said pump; means for determining the enthalpy of said liquid in the inlet of said pump from the pressure and temperature of said liquid; means for comparing the enthalpy of said liquid in said inlet against the required subcooling and for generating a first indicative signal when the enthalpy of said liquid fails to exceed the required subcooling, whereby said first indicative signal indicates potential cavitation in said pump; means for developing a signal indicative of the instantaneous power consumed by said motor; means for comparing the level of power consumption by said motor against a predetermined maximum power and generating a second indicative signal should said power limited be exceeded whereby said second indicative signal indicates potential pump motor overload; an OR GATE for receiving said first and second indicative signals and generating a unipolarity second control signal in response to either said first or said second indicative signals; and control signal summing means for algebraically summing said second control signal with said first control signal to develop a valve position signal to close the positioning of said flow control means; whereby said flow control means is moved toward its closed position in response to an indication of motor overload or potential cavitation in said pump. (a) measuring the pressure of said liquid flowing into said pump and generating a pressure indication signal in response thereto; (b) measuring the temperature of said liquid flowing into said pump and generating a temperature indication signal in response thereto; (c) providing from said temperature a comparison signal related to the required subcooling for said pump; (d) at least periodically developing an enthalpy indication signal correlated with the temperature and pressure indication signals; (e) comparing said required subcooling signal with said enthalpy signal; and (f) actuating said flow control means to steadily reduce liquid flow through said flowline so long as said enthalpy signal fails to exceed required subcooling as determined in step (e). (i) generating an indication signal of saturation pressure as a function of water temperature in said pump inlet, and (ii) subtracting said saturation pressure indication signal from said pressure signal. (g) monitoring the rate of power consumption by said motor; (h) continually comparing the rate of power consumption with a predetermined allowable maximum rate of power consumption; and (i) providing a microprocessor with memory, said memory being adapted to provide subcooling indications for discrete combinations of water temperature and water pressure, (ii) introducing said temperature and pressure indications to said microprocessor whereby said microprocessor is enabled to periodically perform a table look-up operation for said discrete subcooling indication, and (iii) providing means to convert said subcooling indication to an analog indication of subcooling. means for measuring the pressure of said liquid in the inlet of said pump; means for measuring the temperature of said liquid in the inlet of said pump; means for indicating the required subcooling for said liquid entering said pump at the measured temperature of said liquid; means for determining the enthalpy of said liquid in the inlet of said pump from the measured pressure and measured temperature of said liquid; means for comparing the enthalpy of said liquid in said inlet against the required subcooling and for generating a second control signal when the enthalpy of said liquid fails to exceed said required subcooling; and control signal summing means for algebraically subtracting said second control signal from said first control signal to develop a prime mover control signal to control energization of said prime mover; whereby said prime mover is energized at a level such that cavitation in said pump is prevented. 2. In a system as set forth in claim 1, said first indicative signal generating means including limit trigger means effective to generate a fixed potential output signal for said first indicative signal. 3. In a system as set forth in claim 2, signal limiting means for limiting said first control signal to a maximum value when it is of said first polarity. 4. In a system as set forth in claim 3, signal integrating means disposed to increase said second control signal over time before introduction to said summing circuit to insure that said second control signal will dominate said first control signal. 5. In a system as set forth in claim 4, wherein said subcooling determination means includes digital electronics means including a memory, whereby said pressure and temperature indications may be processed to facilitate addressing an appropriate register in said memory to develop subcooling indication. 6. In a system having a flowline for transporting a liquid, a pump with an inlet and an outlet in said flowline, means to drive said pump, and flow control means adapted to control flow through said pump, a method for protecting said pump comprising the steps of: 7. In a method as set forth in claim 6, wherein said step of at least periodically determining said enthalpy indication is provided by; 8. In a method as set forth in claim 7, wherein said step of comprising is done by generating from said temperture indication a signal related to the required minimum pressure difference between actual pump inlet pressure and saturation pressure at current temperature. 9. In a method as set forth with claim 6, wherein said means for driving said pump comprises an induction electric motor, said method including the additional steps of: 10. In a method as set forth in claim 7, wherein said step of at least periodically determining said enthalpy indication is provided by; 11. In a nuclear power plant system having a fluid-filled reactor vessel with a vapor outflow line for removing vapor from said reactor vessel, liquid inflow means for injecting liquid to said reactor vessel, said inflow means including an inflow line, a centrifugal pump disposed along said inflow line, a controllable prime mover for driving said pump, and means for generating a first control signal in response to a liquid level in said vessel and net vapor outflow versus liquid inflow with respect to said vessel, a pump system protection apparatus comprising: 12. In a system as set forth in claim 11, said means for generating a second control signal comprising a trigger signal generator adapted to generate a constant valued output signal for said second control signal. 13. In a system as set forth in claim 12, signal limiting means for limiting said first control signal to a maximum value when it indicates a demand for increased energization of said prime mover. 14. In a system as set forth in claim 13, signal integrating means disposed to increase said second control signal over time, before application to said summing circuit, to insure that said second control signal will dominate said first control signal should both be present. 15. In a system as set forth in claim 1, wherein said working medium is water.
	description	This application claims priority under 35 USC §119 to KR 10-2010-0007938 filed Jan. 28, 2010. The disclosure of which is expressly incorporated by reference herein in their entirety. 1. Field of the Invention The present invention relates to a guide thimble plug for coupling a guide thimble and a shock absorption tube to a bottom nozzle of a nuclear fuel assembly. 2. Description of the Related Art As is well known to those skilled in the art, a nuclear reactor is a device in which a fission chain reaction of fissionable materials is controlled for the purpose of generating heat, producing radioactive isotopes and plutonium, or forming a radiation field. Generally, in light-water reactor nuclear power plants, enriched uranium (U) is used, in which the proportion of U-235 has been increased by 2-5%. To process enriched uranium into nuclear fuel to be used in nuclear reactors, uranium is formed into a cylindrical pellet having a weight of about 5 g. Several hundreds of pellets are retained in a bundle, and are inserted into a zirconium tube under vacuum conditions. A spring and helium gas are placed into the tube, and a cover is welded and sealed onto the tube, thus completing the fuel rod. A plurality of fuel rods constitutes a nuclear fuel assembly and is burned in a nuclear reactor by nuclear reaction. FIG. 1 illustrates such a nuclear fuel assembly and elements thereof. Referring to FIG. 1, the nuclear fuel assembly includes a frame body and a plurality of fuel rods 1. The frame body includes a top nozzle 4, a bottom nozzle 5, a plurality of support grids 2, a plurality of guide thimbles 3 and a measurement tube 6. The fuel rods 1 are inserted through the support grids 2 and supported by springs (not shown) and dimples (not shown) which are formed in the support grids 2. In order to assemble the nuclear fuel assembly, lacquer is applied to the surfaces of the fuel rods 1 to prevent the fuel rods 1 from being scratched, and to prevent the springs provided in the support grids 2 from being damaged. Thereafter, the fuel rods 1 are installed in the frame body, and then the top and bottom nozzles 4 and 5 are coupled to the guide thimbles 3, thus completing the assembly of the nuclear fuel assembly. The assembled nuclear fuel assembly is tested for distances between the fuel rods, distortion, dimensions including the length, etc., after the lacquer is removed, thus completing the process of manufacturing the nuclear fuel assembly. Meanwhile, the guide thimbles 3 provide passages into which control rods (not shown) can be inserted, which are used to operate or stop the nuclear reactor or control the output of the reactor. When it is desired to suddenly stop the nuclear reactor, the control rods free-fall into the guide thimbles 3. Here, to absorb impact generated by free-fall of the control rods, a shock absorption tube is provided in the lower end of each guide thimble 3. As shown in FIG. 2A, the shock absorption tube may be formed by reducing the inner and outer diameters of the lower end of a guide thimble 3′. Alternatively, as shown in FIG. 2B, a separate shock absorption tube 7 having a diameter less than the inner diameter of the guide thimble 3 may be inserted into the guide thimble 3. Recently, to increase lateral resistance and for ease of assembly, a double tube structure like that of FIG. 2B, in which a shock absorption tube is manufactured through a separate process and inserted into the guide thimble, is being used more frequently. Here, in the case of the double tube structure, the guide thimble 3 and the shock absorption tube 7 are coupled to the bottom nozzle 5 by a guide thimble plug C. In detail, as shown in FIG. 3, the guide thimble 3 and the shock absorption tube 7 are coupled to the guide thimble plug C. The bottom nozzle 5 is fastened to the guide thimble plug C by a screw 9. Thereafter, the shock absorption tube 7 is further reliably fastened to the guide thimble 3 by a welding method or a bulging method using plastic deformation of the guide thimble 3 and the shock absorption tube 7. As such, compared to the case of the guide thimble 3′ which is reduced in diameter in the lower end thereof to form the shock absorption tube, the double tube structure including the guide thimble 3 and the shock absorption tube 7 can be more easily manufactured, and resistance with respect to a lateral load can be increased. Thus, the double tube structure 3, 7 has an advantage over the guide thimble 3′ in preventing the nuclear fuel assembly from being bent. Representative examples of the double tube structure of the guide thimble were disclosed in U.S. Pat. No. 4,655,990 entitled “Fuel Assemblies for Nuclear Reactor,” and U.S. Pat. No. 5,068,083 entitled “Dashpot Construction for a Nuclear Reactor Rod Guide Thimble.” However, in conventional techniques, when a single guide thimble plug is welded to the double tube structure, a welded portion may be deformed, resulting in the assembly of the shock absorption tube and the guide thimble being very difficult. Therefore, the quality of the product is diminished. In other words, in the case where the shock absorption tube and the guide thimble which form the double tube structure are assembled with the single guide thimble plug by welding, the welded portion may be deformed. As a result, when the guide thimble is assembled with other elements of the nuclear fuel assembly, a large load is applied to the nuclear fuel assembly, thus reducing the productivity, and reducing the quality of the product. That is, when the upper end of the guide thimble plug is coupled to the shock absorption tube 7, if welding is used as a means for coupling, the straightness of the guide thimble and the shock absorption tube can be compromised, or a welded portion of the shock absorption tube is expanded in diameter so that it becomes very difficult to insert the shock absorption tube into the guide thimble. Furthermore, if the guide thimble plug is coupled to the shock absorption tube by force-fitting, when the guide thimble plug is welded to the guide thimble, the shock absorption tube may be undesirably loosened or removed from the guide thimble plug by welding heat. Accordingly, the present invention has been made keeping in mind the above problems occurring in the prior art, and an object of the present invention is to provide a guide thimble plug for a nuclear fuel assembly which is configured such that a guide thimble and a shock absorption tube which form a double tube structure can be reliably fastened to a bottom nozzle, and thermal deformation of the guide thimble can be minimized. In order to accomplish the above object, the present invention provides a guide thimble plug for coupling a guide thimble having a shock absorption tube therein to a bottom nozzle of a nuclear fuel assembly. The guide thimble plug includes a main body. The main body has: an internal threaded hole formed therethrough so that the main body is coupled to the bottom nozzle by screw coupling; an upper insert part formed in an upper end of the main body, the upper insert part being inserted into the shock absorption tube; and a thermal deformation prevention part formed on the main body below the upper insert part, the thermal deformation prevention part being recessed inwards from an outer surface of the main body such that, when the main body is coupled to the guide thimble, a gap is defined between the thermal deformation prevention part and the guide thimble. Preferably, an external thread can be formed on a circumferential outer surface of the upper insert part, so that the upper insert part is threaded into the shock absorption tube. Alternatively, a caulking groove can be formed on a circumferential outer surface of the upper insert part in a circumferential direction, so that the upper insert part is coupled to the shock absorption tube by caulking. As a further alternative, caulking depressions can be formed on a circumferential outer surface of the upper insert part at positions spaced apart from each other with respect to a circumferential direction, so that the upper insert part is coupled to the shock absorption tube by caulking. The main body can further have a protruding part provided between the upper insert part and the thermal deformation prevention part. The protruding part supports a lower end of the shock absorption tube and has a diameter greater than a diameter of the thermal deformation prevention part, such that the protruding part is forcibly fitted into the guide thimble. Preferably, a width of the thermal deformation prevention part can be two or more times greater than a width of the protruding part. Furthermore, caulking groove indicators can be respectively formed in an upper surface of the protruding part at positions corresponding to the caulking depressions. In the present invention, a shock absorption tube and a guide thimble can be reliably coupled to a bottom nozzle using a single guide thimble plug. Furthermore, thermal strain of the guide thimble can be minimized when welding for coupling the guide thimble plug to the guide thimble. Hence, the efficiency with which the nuclear fuel assembly is assembled and manufactured can be markedly enhanced. Hereinafter, preferred embodiments of a guide thimble plug for a nuclear fuel assembly according to the present invention will be described in detail with reference to the attached drawings. FIG. 4 is a sectional view illustrating a guide thimble plug for a nuclear fuel assembly according to a first embodiment of the present invention. As shown in FIG. 4, the guide thimble plug 30 according to the first embodiment has an approximately cylindrical main body. The guide thimble plug 30 has an upper insert part 32 which is provided on the upper end of the cylindrical main body and is tightened into a shock absorption tube 20, and a thermal deformation prevention part 34 which is recessed from the circumferential outer surface of the approximate medial portion of the guide thimble plug 30, such that a gap is defined between the thermal deformation prevention part 34 and the inner surface of a guide thimble 10 when the guide thimble plug 30 is fitted into the guide thimble 10. An internal threaded hole 31 is formed through the guide thimble plug 30 so that a bottom nozzle is coupled to the guide thimble plug 30 by screw coupling. An external thread 32a is formed on the circumferential outer surface of the upper insert part 32 so that the upper insert part 32 is threaded into the shock absorption tube 20. Furthermore, a protruding part 33 is provided between the upper insert part 32 and the thermal deformation prevention part 34. The protruding part 33 is forcibly fitted into the guide thimble 10. For this, an outer diameter D1 of the protruding part 33 is greater than an inner diameter of the guide thimble 10. An outer diameter D2 of the thermal deformation prevention part 34 is less than the outer diameter D1 of the protruding part 33, so that the force required to fit the guide thimble plug 30 into the guide thimble 10 is slightly mitigated, thus minimizing radial strain on the guide thimble 10 when it is coupled to the guide thimble plug 30. Preferably, a chamfered surface 33a is formed on the upper end of the protruding part 33 to facilitate insertion of the guide thimble plug 30 into the guide thimble 10. The guide thimble 10 and the guide thimble plug 30 are coupled to each other by butt welding on the end of the thermal deformation prevention part 34. If the thermal deformation prevention part 34 is not recessed from the outer surface of the guide thimble plug 30, when the welding is conducted after the thermal deformation prevention part 34 is forcibly fitted into the guide thimble 10, the thermal deformation prevention part 34 is expanded in diameter by thermal strain and the outer diameter thereof is increased. Thus, when the thermal deformation prevention part 34 passes through a sleeve (not shown) of a support grid 2, an excessive load can be applied to the sleeve or the guide thimble 10. However, in the embodiment, because the thermal deformation prevention part 34 is recessed inward from the outer surface of the guide thimble plug 30, thermal strain when welding can be minimized. Here, it is preferable that a width W2 of the thermal deformation prevention part 34 be two or more times greater than a width W1 of the protruding part 33, which is a portion of the guide thimble plug 30 and is substantially forcibly fitted into the guide thimble 10. This mitigates the thermal strain affecting the protruding part 33 which functions to maintain the forcibly fitted state between the guide thimble 10 and the guide thimble plug 30 when welding, thus minimizing the expansion of the protruding part 33 when welding. FIG. 5 is a sectional view illustrating a guide thimble plug for a nuclear fuel assembly according to a second embodiment of the present invention. As shown in FIG. 5, the second embodiment shows another example of the coupling of the guide thimble plug to the shock absorption tube. The guide thimble plug 40 according to the second embodiment includes an internal threaded hole 41, an upper insert part 42, a protruding part 43, a chamfered surface 43a and a thermal deformation prevention part 44, in the same manner as that of the first embodiment. However, unlike the first embodiment in which the external thread 32a is formed on the circumferential outer surface of the upper insert part 32, in the second embodiment, a caulking groove 42a is formed in a circumferential direction on the outer surface of the upper insert part 42 of the guide thimble plug 40 which is fitted into a shock absorption tube 20. After the upper insert part 42 of the guide thimble plug 40 is inserted into the shock absorption tube 20, several striking points are set on the surface of the shock absorption tube 20 at positions corresponding to the caulking groove 42a of the guide thimble plug 40, and impact is applied to the striking points. Then, the striking points of the shock absorption tube 20 are fitted into the caulking groove 42a, so that the shock absorption tube 20 can be reliably united with the upper insert part 42 of the guide thimble plug 40. FIG. 6 illustrates a guide thimble plug 50 according to a third embodiment of the present invention. As shown in FIG. 6, the guide thimble plug 50 according to the third embodiment includes an internal threaded hole 41, an upper insert part 42, a protruding part 43, a chamfered surface 43a and a thermal deformation prevention part 44 in the same manner as the guide thimble plug 40 of the second embodiment. In the case of the third embodiment, caulking depressions 52a are formed on the outer circumference of the upper insert part 42 at positions spaced apart from each other at predetermined intervals with respect to the circumferential direction, unlike the second embodiment in which the caulking groove 42a is formed in the circumferential direction on the outer surface of the upper insert part 42. Caulking groove indicators 53a are formed in the upper surface of the protruding part 43 at positions corresponding to the lower ends of the relative caulking depressions 52a. Due to the caulking groove indicators 53a, even after a shock absorption tube 20 is fitted over the guide thimble plug 50, the positions of the caulking depressions 52a can be easily indicated. Therefore, after the upper insert part 42 of the guide thimble plug 50 is inserted into the shock absorption tube 20, striking points are set on the surface of the shock absorption tube 20 at positions corresponding to the caulking depressions 52a using the caulking groove indicators 53a, and impact is applied to the striking points. Then, the striking points of the shock absorption tube 20 are fitted into the caulking depressions 52a so that the shock absorption tube 20 can be reliably united with the upper insert part 42 of the guide thimble plug 50. In this case, because the caulking depressions 52a are formed at positions spaced apart from each other in the circumferential direction, the shock absorption tube 20 and the guide thimble plug 50 which are coupled to each other can be prevented from undesirably rotating with respect to each other around the longitudinal central axis of the shock absorption tube 20. Instead of the caulking, without the caulking depressions 52a, spot welding can be utilized to assemble the shock absorption tube 20 and the guide thimble plug 50. As such, according to the first through third embodiments of the present invention, the upper insert part of the guide thimble plug is united with the shock absorption tube by screw coupling or caulking, and the guide thimble plug is coupled to the guide thimble by welding. Therefore, the shock absorption tube and the guide thimble can be reliably coupled to the bottom nozzle using the single guide thimble plug. Furthermore, because the guide thimble plug has the thermal deformation prevention part, thermal strain on the guide thimble can be minimized when welding for coupling the guide thimble plug to the guide thimble. With the present invention, the efficiency with which the nuclear fuel assembly is assembled and manufactured can be markedly enhanced. Although the preferred embodiments of the present invention have been disclosed for illustrative purposes, those skilled in the art will appreciate that various modifications, additions and substitutions are possible, without departing from the scope and spirit of the invention as disclosed in the accompanying claims.
056404344	claims	1. A miniaturized nuclear reactor utilizing improved pressure tube structural member comprising: a) a moderator (20) having: a top reactor wall, a bottom reactor wall, a reactor front wall, a reactor back wall, and two reactor side walls; b) at least one calandria tube (12) contained within the moderator (20); c) at least one fuel channel pressure tube (14) contained within the at least one calandria tube (12), the fuel channel pressure tube (14) further comprises: a fuel channel pressure tube coating (14A), the at least one fuel channel pressure tube (14) further comprises a fuel channel pressure tube lining (14B) and a fuel channel pressure tube cladding (14C); d) at least one fuel bundle (26) contained within the at least one fuel channel pressure tube (14); e) at least one fuel channel pressure tube pad (18) positioned between the at least one calandria tube (12) and the at least one fuel channel pressure tube (14), the fuel channel pressure tube pad (18) further comprises a pair of fuel channel pressure tube pad vertical spacers (18A) which are at an obtuse angle to each of a pair of fuel channel pressure tube pad ends (18B), the fuel channel pressure tube pad (18) further comprises a fuel channel pressure tube pad horizontal spacer (18C) connected at opposite distal ends to each one of the fuel channel pressure tube pad vertical spacers (18A); f) at least one horizontal exterior support pad (30) positioned on the bottom reactor wall, the at least one horizontal exterior support pad (30) is positioned between the at least one calandria tube (12) and the bottom reactor wall; g) at least one fuel bundle support pad (16) positioned between the at least one fuel bundle (26) and the at least one fuel channel pressure tube (14); h) at least one vertical support pad (24) positioned on a reactor side wall between the at least one calandria tube (12) and the reactor side wall; I) at least one angular support pad (28) positioned on a corner formed between the reactor top wall and the reactor side wall, the at least one angular support pad (28) extending angularly from the corner, the at least one angular support pad functioning to maintain a space between the at least one calandria tube (12) and the reactor top and side walls; j) at least one moderator system comprises: at least one moderator inlet, at least one moderator outlet, at least one moderator pump, and at least one moderator cooler; k) at least one coolant system which comprises: at least one coolant inlet, at least one coolant outlet, and at least one coolant gas fan; l) at least one heat exchanger system which comprises: at least one inlet, at least one outlet, at least one turbine, at least one generator, at least one condenser, and at least one feed water pump. a) a moderator (20) having: a top reactor wall, a bottom reactor wall, a reactor front wall, a reactor back wall, two reactor side walls, and a reactor horizontal interior wall; b) a pair of calandria tubes (12) which comprise an upper calandria tube and a lower calandria tube contained within the moderator (20); c) a pair of fuel channel pressure tubes (14) which comprise an upper fuel channel pressure tube contained within the upper calandria tube and a lower channel pressure tube contained within the lower calandria tube; d) a pair of fuel bundles (26) which comprise an upper fuel bundle contained within the upper fuel channel pressure tube and a lower fuel bundle contained within the lower fuel channel pressure tube; e) at least two fuel channel pressure tube pads (18) which comprise an upper fuel channel pressure tube pad positioned between the upper calandria tube and the upper fuel channel pressure tube and a lower fuel channel pressure tube pad positioned between the lower calandria tube and the lower fuel channel pressure tube; f) at least one horizontal exterior support pad (30) positioned on the bottom reactor wall, the at least one horizontal exterior support pad (30) is positioned between the lower calandria tube and the bottom reactor wall; g) at least two fuel bundle support pads (16) which comprise an upper fuel bundle support pad positioned between the upper fuel bundle and the upper fuel channel pressure tube and a lower fuel bundle support pad positioned between the lower fuel bundle and the lower fuel channel pressure tube; h) at least two vertical support pads (24) which comprise an upper vertical support pad positioned on an upper reactor side wall between the upper calandria tube and the upper reactor side wall and a lower vertical support pad positioned on a lower reactor side wall between the lower calandria tube and the lower reactor side wall; I) at least two angular support pads (28) which comprise an upper angular support pad positioned on a corner formed between the reactor top wall and the reactor side wall the upper angular support pad extending angularly from the corner and a lower angular support pad positioned on a corner formed between the reactor bottom wall and the reactor side wall, the lower angular support pad extending angularly from the corner, the upper angular support pad functioning to maintain a space between the upper calandria tube and the reactor top and side walls, the lower angular support pad functioning to maintain a space between the lower calandria tube and the reactor bottom and side walls; j) at least one horizontal interior support pad (22) positioned between an upper calandria tube and the at least one reactor horizontal interior wall; k) at least one moderator system comprises: at least one moderator inlet, at least one moderator outlet, at least one moderator pump, and at least one moderator cooler; l) at least one coolant system which comprises: at least one coolant inlet, at least one coolant outlet, and at least one coolant gas fan; and m) at least one heat exchanger system which comprises: at least one inlet, at least one outlet, at least one turbine, at least one generator, at least one condenser, and at least one feed water pump. A) a first fuel bundle proximal end plate (40AA) having a plurality of first fuel bundle proximal end plate fuel element end fasteners (40AAA), further having a plurality of first fuel bundle proximal end plate ports (40AAB), still further having a first fuel bundle proximal end plate indent (40AAC), yet still further having a first fuel bundle proximal end plate opening (40AAD); B) a second fuel bundle distal end plate (40BA) having a plurality of second fuel bundle distal end plate fuel element end fasteners (40BAA), a plurality of second fuel bundle distal end plate ports (40BAB), a second fuel bundle distal end plate indent (40BAC) and a second fuel bundle distal end plate opening (40BAD); C) a fuel bundle support (40D) having a fuel bundle support proximal end (40DA), a fuel bundle support proximal end spacer (40DB), a fuel bundle support distal end (40DC), a fuel bundle support distal end spacer (40DD), a fuel bundle support spacer tube (40DE), a fuel bundle support rod (40DF), and a fuel bundle support nut (40DG); and D) a plurality of fuel elements (40C) each having a proximal and distal end, the proximal end of each fuel element (40C) affixed within each of the first fuel bundle proximal end plate fuel element end fasteners (40AAA) and the distal end of each fuel element (40C) affixed within each of the second fuel bundle distal end plate fuel element end fasteners (40BAA). a) a moderator (20) having: a left top reactor wall, a right top reactor wall, a bottom left reactor wall, a bottom right reactor wall, a reactor front wall, a reactor back wall, a reactor upper left exterior side wall, a reactor lower left exterior side wall, a reactor upper right exterior side wall, a reactor lower right exterior side wall, at least one reactor horizontal interior wall having a reactor horizontal interior left and right wall, and at least one reactor vertical interior wall having a reactor vertical interior upper and lower wall, the moderator (20) divided into at least four compartments which comprise an upper left compartment, an upper right compartment, a lower left compartment, and a lower right compartment, the upper left compartment being bordered at the bottom by a reactor left horizontal interior wall and a reactor upper vertical interior wall and a reactor upper left exterior side wall and a reactor left top wall, the upper right compartment is bordered at the bottom by a reactor right horizontal interior wall and a reactor upper vertical interior wall and a reactor upper right exterior side wall and a reactor upper right top wall, the lower left compartment is bordered at the bottom by a reactor left bottom wall and a reactor left horizontal interior wall and a reactor lower vertical interior wall and a reactor lower left exterior side wall and a reactor left bottom wall, the lower right compartment is bordered at the bottom by a reactor right bottom wall and a reactor right horizontal interior wall and a reactor lower right exterior side wall and a reactor right bottom wall; b) at least four calandria tubes (12) which comprise an upper left calandria tube, upper right calandria tube, a lower left calandria tube, and a lower right calandria tube contained within the upper left, upper right, lower left, and lower right compartments, respectively, of the moderator (20); c) at least four fuel channel pressure tubes (14) each fuel channel pressure tube contained within each of the calandria tubes; d) at least four fuel bundles (26) each fuel bundle contained within each of the fuel channel pressure tubes; e) at least four fuel channel pressure tube pads (18) each fuel channel pressure tube pad positioned between each of the calandria tubes and each of the fuel channel pressure tubes; f) at least two horizontal exterior support pad (30), one horizontal exterior support pad positioned on the left bottom reactor wall and the other horizontal exterior support pad positioned on the right bottom reactor wall, the horizontal exterior support pads (30) are positioned between the lower left and lower right calandria tubes and the bottom left and bottom right reactor walls, respectively; g) at least four fuel bundle support pads (16), each bundle support pad positioned between each of the fuel bundles and each of the fuel channel pressure tubes; h) at least two vertical support pads (24) which comprise an upper vertical support pad positioned on an upper reactor vertical interior wall between the upper left and upper right calandria tubes and a lower vertical support pad positioned on a lower reactor side wall between the lower left and lower right calandria tubes; I) at least four angular support pads (28) which comprise an upper left angular support pad positioned on a corner formed between the reactor left top wall and the reactor upper left exterior side wall, the upper left angular support pad extending angularly from the corner and an upper right angular support pad positioned on a corner formed between the reactor right top wall and the reactor upper right exterior side wall, the upper right angular support pad extending angularly from the corner, and a lower left angular support pad positioned on a corner formed between the reactor left bottom wall and the reactor lower left exterior side wall, the lower left angular support pad extending angularly from the corner, and a lower right angular support pad positioned on a corner formed between the reactor right bottom wall and the reactor lower right exterior side wall, the lower right angular support pad extending angularly from the corner, the angular support pads functioning to maintain a space between the calandria tube and the reactor walls; j) at least two horizontal interior support pads (22), one horizontal interior support pad positioned within the left reactor horizontal interior wall and the other horizontal interior support pad positioned within the right reactor horizontal interior wall, the left horizontal interior support pad functioning to maintain a space between the upper left and lower left calandria tubes, and right horizontal interior support pad functioning to maintain a space between the upper right and lower right calandria tubes; k) at least one moderator system which comprises: at least one moderator inlet, at least one moderator outlet, at least one moderator pump, and at least one moderator cooler; l) at least one coolant system which comprises: at least one coolant inlet, at least one coolant outlet, and at least one coolant gas fan; and m) at least one heat exchanger system which comprises: at least one inlet, at least one outlet, at least one turbine, at least one generator, at least one condenser, and at least one feed water pump. a) a moderator (20) having: a top reactor wall, a bottom reactor wall, a reactor front wall, a reactor back wall, and two reactor side walls: b) at least one second calandria tube (112) having a plurality of second calandria tube compartments (112A) securely affixed around an inside perimeter contained within the moderator (20); c) at least one second fuel channel pressure tube (114) having a plurality of second fuel channel pressure tube compartments (114A) securely affixed around an outside perimeter, the plurality of second calandria tube compartments (112A) and the plurality of second fuel channel pressure tube compartments (114A) opposing each other in an interlocking configuration, the at least one second fuel channel pressure tube (114) is contained within the at least one second calandria tube (112), d) at least one fuel bundle (26) contained within the at least one second fuel channel pressure tube (114); e) at least one second fuel channel pressure tube support pad (113) having a second fuel channel pressure tube support pad end (113A), a second fuel channel pressure tube support pad spacer (113B), a second fuel channel pressure tube support pad concave (113C), a second fuel channel pressure tube support pad convex (113D), a second fuel channel pressure tube support pad groove (113E), and a second fuel channel pressure tube support pad opening (113F) positioned between the at least one calandria tube (12) and the at least one second fuel channel pressure tube (114); f) at least one horizontal exterior support pad (30) positioned on the bottom reactor wall, the at least one horizontal exterior support pad (30) is positioned between the at least one second calandria tube (112) and the bottom reactor wall; g) at least one fuel bundle support pad (16) positioned between the at least one fuel bundle (26) and the at least one fuel channel pressure tube (14); h) at least one vertical support pad (24) positioned on a reactor side wall between the at least one calandria tube (12) and the reactor side wall; i) at least one angular support pad (28) positioned on a corner formed between the reactor top wall and the reactor side wall, the at least one angular support pad (28) extending angularly from the corner, the at least one angular support pad functioning to maintain a space between the at least one calandria tube (12) and the reactor top and side walls; j) at least one moderator system comprises: at least one moderator inlet, at least one moderator outlet, at least one moderator pump, and at least one moderator cooler; k) at least one coolant system which comprises: at least one coolant inlet, at least one coolant outlet, and at least one coolant gas fan; and l) at least one heat exchanger system which comprises: at least one inlet, at least one outlet, at least one turbine, at least one generator, at least one condenser, and at least one feed water pump. 2. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 1, wherein the moderator (20) further comprises: 3. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 1, wherein the calandria tube (12) further comprises: a calandria tube coating (12A); a calandria tube lining (12B); and a calandria tube cladding 12C. 4. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 1, wherein the fuel bundle (16) further comprises a pair of fuel bundle support pad spacers (16A) positioned at opposite distal ends of a fuel bundle support pad strap (16B). 5. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 1, wherein the at least one fuel bundle (26) is manufactured from glass. 6. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 2, wherein the horizontal interior support pad further comprises: a horizontal interior support pad proximal end (22A), a horizontal interior support pad distal end (22B), a horizontal interior support pad groove (22C), and a horizontal interior support pad concave (22D), the horizontal interior support pad groove (22C) fitting snugly around a reactor horizontal interior wall, the horizontal interior support pad concave (22D) functioning as a cradle upon which the upper calandria tube rests upon. 7. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 6, wherein the horizontal interior support pad further comprises: a horizontal interior support pad coating (22E), a horizontal interior support pad lining (22F), and a horizontal interior support pad cladding (22G). 8. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 1, wherein the angular support pad (28) further comprises: an angular support pad top member (28A) and an angular support pad bottom member (28B), the angular support pad top member (28A) abutting the calandria tube (12), and the angular support pad bottom member (28B) securely fastened to the corner. 9. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 1, wherein the horizontal exterior support pad (30) further comprises a horizontal exterior support pad end (30A), a horizontal exterior support pad fastener (30B), and a horizontal exterior support pad concave (30C), the horizontal exterior support pad fastener (30B) securely affixes the horizontal exterior support pad (30) to the reactor bottom wall, and the horizontal exterior support pad concave (30C) functioning as a cradle within which the calandria tube (12) rests. 10. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 1, wherein the fuel bundle (40) further comprises: 11. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 1, wherein the moderator (20) further comprises: 12. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 11, wherein the vertical support pad (24) further comprises at least two vertical support pads (24) which comprise an upper vertical support pad and a lower vertical support pad, the upper and lower vertical support pads are securely affixed within the reactor upper interior vertical wall and the reactor lower interior wall, respectively, each of the vertical support pads (24) having a vertical support pad proximal end (24A), a vertical support pad distal end (24B), a vertical support pad groove (24C), and a vertical support pad concave (24D), the vertical support pad groove (24C) fitting snugly around a upper reactor vertical interior wall and a lower reactor vertical interior wall, respectively, the vertical support pad concave (24D) of the upper vertical support pad functioning as a cradle upon which the interior sides of the upper left calandria tube and the upper right calandria tube rests upon, the lower vertical support pad functioning as a cradle upon which the interior sides of the lower left calandria tube and the lower right calandria tube rests upon. 13. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 1, wherein the reactor further comprises: a least one joint connector (48) having a joint connector tread (48A), a joiner ring (50) with a joiner ring thread (50), the joint connector tread (48A) connects to the joiner ring thread (50) and a service tube (52) connected to the joiner ring (50) functioning for refueling of the reactor. 14. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 2, wherein the calandria tubes further comprise a fuel compartment pressure tube (17) therebetween. 15. A miniaturized nuclear reactor utilizing improved pressure tube structural members comprising: 16. The miniaturized nuclear reactor utilizing improved pressure tube structural members as described in claim 15, wherein the at least one fuel bundle (26) is manufactured from glass.
055815871	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS Preferred embodiments of the present invention will be described hereunder with reference to the accompanying drawings. A first embodiment of a control rod driving apparatus according to the present invention will now be described with reference to FIG. 1. Referring to FIG. 1, the same elements as those shown in FIG. 10 are given the same reference numerals and their descriptions are omitted here. A hydraulic motor is accommodated in a motor case 52, the hydraulic motor comprising a hydraulic motor 53 for insertion, a hydraulic motor 54 for withdrawal, bevel gears 55a and 55b serving as transmission mechanisms to convert the direction of rotations of a driving shaft of the hydraulic motor so as to transmit the power to the ball spindle 5 and worm gears 56a and 56b. For the hydraulic motor, one of a gear motor, a vane motor, a piston motor, a turbine motor and so forth may be employed. In this embodiment, the maintenance facility is improved by using a turbine-type hydraulic motor because it has a simple structure composed of small number of sliding elements and therefore its periodical maintenance labor requirements can be decreased. If the normal revolving speed of the turbine-type hydraulic motor is higher than the revolving speed of the ball spindle 5 which is required to drive the control rod 11 in a steady state, the revolving speed is lowered by the worm gears 56a and 56b and the bevel gears 55a and 55b. An embodiment in a case where the operational speed of the transmission mechanism is lowered will now be described. Since the turbine-type hydraulic motor is generally designed in such a manner that its efficiency is improved when it is rotated forwards or reverse, this embodiment comprises a hydraulic motor 53 for insertion and hydraulic motor 54 for withdrawal which are used as the hydraulic motors to respectively insert and withdraw the control rod. When hydraulic pressure is applied to the hydraulic motor 53 for insertion for example, a rotational force is generated in a predetermined direction, the rotational force being transmitted from the bevel gear 55a supported by bearings 62 to the ball spindle 5 through the worm gear 56a disposed coaxially with the bevel gear 55b and the worm gear 56b supported by bearings 63. When the ball spindle 5 is rotated, the ball nut 9 allowed to engage with the ball spindle 5 is moved upwards. Simultaneously, the connection pipe 10 placed on the ball nut 9 is moved upwards. Therefore, the control rod 11 is inserted into the reactor core. Since the turbine-type hydraulic motor has substantially no driven torque in general, another mechanism is required to maintain the positions of the control rod 11 and the connection pipe 10. In this embodiment, the engagement operation of the worm gear 56 maintains the positions of the control rod 11 and the connection pipe 10 even if a vertical force is applied to the same. Therefore, the electromagnetic brake required for the conventional control rod driving apparatus can be omitted from the structure. It is preferable that the number of pipes for supplying hydraulic pressure to the hydraulic motor from the outside of the control rod driving apparatus is minimized in order to simplify the layout of pipes in the lower portion of the pressure vessel for the reactor. In this embodiment, the hydraulic pressure supplied from a hydraulic pipe 57 for driving is changed over by an electromagnetic valve 59 so that the hydraulic pressure is supplied to the hydraulic motor 53 for input and to the hydraulic motor 54 for output through a usual input port 68 and a usual outlet port 64. In this embodiment, discharged water after the hydraulic pressure from the hydraulic motor 53 for input and the hydraulic motor 54 for output has been supplied is discharged into the reactor pressure vessel 2 through a discharge pipe 65. Water may be discharged by another method using a discharge pipe, not shown, communicated with the outside portion of the reactor pressure vessel. If the discharge pipe is provided, maintaining the pressure of the discharged water at a level lower than the pressure in the reactor enables the pressure in the reactor to be used as hydraulic pressure to drive the hydraulic motor. Therefore, the necessity of individually providing a source for driving the hydraulic motor on the outside of the control rod driving apparatus can be eliminated. At the time of the scram, the hydraulic pressure is supplied to a hydraulic pipe 66 for scram operation so that the hydraulic pressure pushes up the connection pipe 10 mounted on the ball nut 9 in the guide tube 8 similarly to the conventional example. As a result, the connection pipe 10 is separated from the ball nut 9 so that scram operation is performed. Although this embodiment has an arrangement such that the water supply at the time of scram operation is performed by means of the hydraulic pipe 66 for scram operation provided individually from the usual driving hydraulic pipe 57, the pipes may be used commonly by employing a structure in which the water passage is changed over by an electromagnetic valve or the like. A second embodiment of the control rod driving apparatus will now be described with reference to FIGS. 2 and 3. Referring to FIG. 2, the same elements as those shown in FIG. 1 are given the same reference numerals and their descriptions are omitted here. The first embodiment has the arrangement in which the pipes for supplying hydraulic pressure for input/output at the time of usual operation of the control rod driving apparatus are formed into a common hydraulic pipe 57 for the usual operation and the electromagnetic valve 59 is used to change over the water passage downstream from the hydraulic pipe 57 for the usual operation. In this second embodiment, the pipes for supplying hydraulic pressure for the insertion operation at the time of the usual operation and those for scram operation are made so that a common introduction port 67 and a change-over valve 60 is disposed to change over the water passage downstream from the introduction port 67. The change-over valve 60 automatically opens/closes the valve in accordance with the flow quantity and the level of the hydraulic pressure so as to change over the water passage for the insertion operation and that for scram operation. FIG. 3 is a schematic view which illustrates the inside portion of the switch valve 60. The change-over valve 60 accommodates valves 68 and 69 which are respectively pressed against valve seats 72 and 73 by springs 70 and 71, respectively. The spring force of the spring 70 is determined so as to be smaller than that of the spring 71. If the hydraulic pressure acting on the introduction port 67 is low, only the valve 68 overcomes the spring force and is separated from the valve seat 72. As a result, the introduction port 67 and the usual insertion port 58 are communicated with each other. The insertion port 58 is communicated with the hydraulic motor 53 for insertion as shown in FIG. 2. Thus, usual insertion is performed similarly to that of the first embodiment. If the hydraulic pressure of the introduction port 67 shown in FIG. 3 is further raised, the valve 68 is pressed against the valve seat 75 by the hydraulic pressure so that communication between the introduction port 67 and the usual insertion port 58 is canceled. Furthermore, the valve 69 overcomes the spring force so as to be separated from the valve seat 73 so that the scram port 80 and the introduction port 67 are communicated with each other. The hydraulic pressure is supplied from the scram port 80 into the guide tube 8 shown in FIG. 2 so that scram is performed. The withdrawal of the control rod 11 shown in FIG. 2 is performed by supplying the hydraulic pressure to a withdrawing pipe 82 to activate the hydraulic motor 54 for withdrawal. In this embodiment, the pipe for supplying the hydraulic pressure for the insertion operation and that for the scram are formed commonly into the introduction port 67. It might be considered feasible to employ a contrary structure in which the pipe for supplying the hydraulic pressure for the withdrawal operation and that for scram are commonly formed. The foregoing case is undesirable because withdrawal operation is performed if the hydraulic pressure of water to be supplied at the time of scram is low. A third embodiment of the control rod driving apparatus according to the present invention will now be described with reference to FIG. 4. Referring to FIG. 4, the same elements as those shown in FIG. 1 are given by the same reference numerals and their descriptions are omitted here. This embodiment employs a turbine-type hydraulic motor 83 which can be rotated both forwards and reversely so that the number of the hydraulic motors is decreased to one. Both insertion and withdrawal of the control rod are performed by the turbine-type hydraulic motor 83. As for the scram, a similar arrangement is made to that according to the first embodiment. A fourth embodiment of the control rod driving apparatus according to the present invention will now be described with reference to FIG. 5. Referring to FIG. 5, the same elements as those shown in FIG. 1 are given the same reference numerals and their descriptions are omitted here. This embodiment has an arrangement that the hydraulic pipes are formed into a multi-pipe structure to decrease the number of the hydraulic pipes so as to simplify the layout of the pipes. The structure shown in FIG. 5 is an example having an arrangement in which a pipe 74 connected to the hydraulic motor 53 for insertion is accommodated in a pipe 85 so that a double-pipe structure is formed. A water flow through a gap between the pipe 85 and the pipe 84 flows through a pipe 86 to be introduced into the hydraulic motor 54 for withdrawal so that the hydraulic motor 54 for withdrawal is operated. Scram operation is performed by supplying hydraulic pressure to the hydraulic pipe 66. The pipe for the scram operation and the pipe for the usual insertion may be formed into a multi-layer pipe to decrease the number of the pipes. A fifth embodiment of the control rod driving apparatus according to the present invention will now be described with reference to FIG. 6. In this embodiment, the change-over valve accommodated in the motor case 52 in the first and third embodiments is disposed on the exterior of the motor case 52. A change-over valve 89 may be disposed at an arbitrary position, such as a position on the inside or outside of the pedestal in the lower portion of the reactor pressure vessel or on the outside of the reactor pressure vessel. The hydraulic pressure supplied through a pipe 90 is, by the change-over valve 89, connected to a usual outlet port 87 or a usual inlet port 88 in accordance with the drive mode so as to operate the hydraulic motor 53 for insertion or the hydraulic motor 54 for withdrawal. In this embodiment, the number of pipes in the pedestal in the lower portion of the pressure vessel for the reactor is increased, resulting in an undesirable pipe layout. However, the ease of maintenance of the change-over valve 89 can be improved. Further, the change-over valve according to the second embodiment may be disposed on the outside of the motor case 52. A sixth embodiment of the control rod driving apparatus according to the present invention will now be described with reference to FIG. 7. The control rod driving apparatus 1 is accommodated in the housing formed integrally with the pressure vessel 2 by welding. A first connection member 93 is attached to the upper portion of a drive shaft 92 which is rotated by a hydraulic motor 91. The drive shaft 92 penetrates a pressure boundary but has no shaft sealing packing. A second connection member 94 is attached to the first connection member 93. A long spindle 95 is connected to the upper portion of the second connection member 94, the spindle 95 having a bearing 96 attached to the top end thereof. The spindle 95 is rotated in synchronization with the drive shaft 92 when the hydraulic motor 91 is rotated. A nut 97 is engaged to the spindle 95, the nut 97 having the top surface with which the lower end of a connection pipe 98 is in contact. Four rollers 99 equally disposed in the circumferential direction are provided for each of the nut 97 and the connection pipe 98, the rollers 99 being in contact with the inner surface of a guide tube 95a. The side surfaces of the rollers 99 are guided by a plate 96b attached to the inner surface of the guide tube 95a to inhibit rotation of the nut 97 and the connection pipe 98. Reference numeral 97b represents a housing. In the control rod drive apparatus 1 constituted as described above, when the spindle 95 is rotated due to the rotations of the hydraulic motor 91, the nut 97 allowed to engage with the spindle 95 is permitted to move only in the axial direction. Therefore, the connection pipe 98 mounted on the nut 97 follows the movement of the nut 97, also causing the control rod 11 connected to the connection pipe 97 to be moved vertically. If the control rod 11 is rapidly inserted, called a scram operation, during an emergency for the reactor, high-pressure water accumulated in an accumulator is supplied through the scram-water injection port 98a to be introduced into the guide tube 95a. As a result, high-pressure water acts on the connection pipe 98 to rapidly pushes the connection pipe 98 in the upward direction. Therefore, the control rod 11 is rapidly inserted into the reactor core. A control rod driving system including a plurality of control rod driving apparatus of the structures described above will be preferably accommodated in a boiling water reactor BWR according to the present invention, which will be described hereunder. FIG. 8 illustrates the structure of a reactor core comprising a plurality of power-adjustment units 150 as designated by diagonal lines. As a control rod driving apparatus for the power-adjustment units 150, any one of the apparatuses according to the first to sixth embodiments is employed. Since the control-rod drive apparatus 1 having the foregoing drive structure causes the control rod 11 to move upwards or downwards in accordance with the rotational angle of the spindle 5 or 95, the control rod 11 can be precisely moved in the core by controlling the rotational angle of the spindle 5 or 95. Therefore, its structure is suitable as a method for driving the control rods for the power adjustment units. The control rod driving apparatus for units except the power adjustment units comprises the control rod driving apparatus having the hydraulic piston drive structure shown in FIG. 11. An example of a system for supplying hydraulic pressure for driving the foregoing control rod driving apparatus will now be described. Among the hydraulic pressure supply systems according to the present invention, the control rod driving apparatus having the hydraulic piston drive structure adapted to units except the power adjustment units 150 comprises the conventional hydraulic pressure supply system shown in FIG. 12. In this case, the conventional structure comprises one stabilizing circuit 105, causing the control rods to be driven one by one. In this embodiment, a plurality of the stabilizing circuits 105 may be provided to enable a plurality of control rods to be driven simultaneously. FIG. 9 illustrates an embodiment of a system for supplying hydraulic pressure to a control rod driving apparatus of a screw-drive type hydraulic drive motor for use in the power adjustment units. It should be noted that a pump 201 of a hydraulic supply portion 200 shown in FIG. 9 may be commonly used with the pump 101 shown in FIG. 12. The hydraulic pressure supply portion 200 has a pipe structure comprising a pump 201, a flow meter 202, a flow-rate adjustment valve 203, a pressure-adjustment valve 204 and a plurality of stabilizing circuits 205. Each stabilizing circuit 205 comprises two systems of electromagnetic valves 206 and 207. One hydraulic-pressure supply portion 200 is provided for one nuclear reactor plant. Pipes represented by pipes 209, 210 and 211 are connected from the hydraulic-pressure supply portion 200 to a hydraulic-pressure control unit 208 which has pipes corresponding to those in the control rod driving apparatus 1. Water flows in the hydraulic-pressure supply portion 200 and in each pipe are designated by arrows. The pipe 209 is a charging pipe for an accumulator 213 which acts when the control rod is inserted to cope with an emergency so that the accumulator 213 is charged with high-pressure water. The accumulator 213 includes a piston 214. The lower portion of the piston 214 is connected to a nitrogen container 216 through a pipe 215. High-pressure nitrogen gas is enclosed in the nitrogen container 216. Reference numeral 217 represents a scram valve which is closed in a usual state so that the accumulator 213 is maintained at a high pressure state. In response to a control-rod emergency insertion signal, the scram valve 217 is opened so that the high-pressure water in the accumulator 213 flows through a scram pipe 219 connected to the lower surface of a connection pipe of the control rod driving apparatus 1 so as to flow in the control rod driving apparatus 1. As a result, a control rod is inserted into the reactor core to cope with emergency. It should be noted that the control rod driving apparatus 1 of the screw-drive structure does not involve scram discharge water. The pipe 210 is a pipe for supplying water for driving the control rod when the output from a reactor is adjusted, the pipe 210 being connected to a direction--control circuit 226 composed of two electromagnetic valves 222 and 223 disposed in the hydraulic-pressure control unit 208. The direction-control circuit 226 acts to change over the rotational direction of the hydraulic drive motor in accordance with the insertion/withdrawal of the control rod. That is, the control rod is inserted by opening the electromagnetic valve 222 so that the driving water flows through the electromagnetic valve 222 and an insertion pipe 220. As a result, the hydraulic drive motor is rotated in a direction which causes the control rod to be inserted. When the control rod is withdrawn, the electromagnetic valve 223 is opened. The driving water flows through the electromagnetic valve 223 and the withdrawing pipe 221 so as to rotate the hydraulic drive motor in a direction which causes the control rod to be withdrawn. In both insertion and withdrawal, drive water rotates the hydraulic drive motor and flows in the pressure vessel 2 through the control-rod drive apparatus 1. The pipe 211 is a purge-water pipe for preventing invasion of foreign materials from the inside portion of the pressure vessel 2 into the control rod driving apparatus 1 so that purge water, the pressure of which has been adjusted to a predetermined level in the purge-water pipe, always flow through the scram pipe 219 to flow in the control rod driving apparatus 1. The electromagnetic valves 206 and 207 of the stabilizing circuit 205 are opened in a usual state so that a quantity required for the insertion of the control rod flows through the electromagnetic valve 206 and a quantity required for the withdrawal of the same flows through the electromagnetic valve 207. As a result, water flows in a purge-water header 227 as a portion of purging water. In the stabilizing circuit 205, the electromagnetic valve 206 is closed when the control rod is inserted to adjust the output so that a quantity of water, which is the same as the quantity of water flowing through the electromagnetic valve 206, flows to the hydraulic drive motor. When the control rod is withdrawn, the electromagnetic valve 207 is closed so that a quantity of water, which is the same as the quantity of water flowing through the electromagnetic valve 207, flows to the hydraulic drive motor. Thus, the stabilizing circuit 205 stabilizes the pressure of drive water similarly to that of the conventional example. Furthermore, a plurality of the stabilizing circuits 205 are provided, thereby simultaneously and stably driving a plurality of control rods. By simultaneously driving control rods of the power adjustment units located at symmetric positions with respect to the central unit in the core for example, the power from the core can be adjusted in such a manner that the symmetry of the distribution of the powers from the core is maintained. By providing a plurality of the stabilizing circuit 205 for the control rod driving apparatus having the hydraulic piston structure as described above, the control rods can be driven similarly in such a manner that the symmetry of the distribution of the powers is maintained. As described above, the control rod driving apparatus according to the present invention comprises a motor which is driven by hydraulic pressure in place of a conventional electric motor so that the necessity of using the shaft sealing packing for the ball spindle can be eliminated and elements which must be periodically changed can be omitted. Therefore, the amount of maintenance required for the control rod driving apparatus can be significantly reduced. As a result, a reduction in the quantity of exposure for operators when a reactor is periodically inspected can be achieved. In addition, it contributes to shorten the time required to complete the latter period of the periodical inspection. Since the shaft sealing portion is omitted from the structure, the drive torque can be reduced and normal operation can be always expected. In addition, the possibility of discharge of the reactor water to the outside of the reactor can be eliminated. As a result, a great contribution can be made to improve the reliability and safety of the reactor. Since the operation of the worm gear maintains the position of the control rod even if external force for vertically moving the control rod acts, the electromagnetic brake, which has been used to prevent the withdrawal of the control rod when a pipe has been broken, can be omitted from the structure. The foregoing boiling water reactor BWR equipped with the control rod driving system according to the present invention enables the control rod to be operated to be adaptable to the function of the core. Therefore, a great contribution can be made to improve controllability of the BWR. By using the motor driven by hydraulic pressure in the control rod driving apparatus of the output adjustment units in place of the conventional electric motor, the necessity of using the shaft sealing packing for the spindle can be eliminated. In addition, the same hydraulic pressure supply system as that for the control rod driving apparatus, except for the power adjustment units, is employed. Therefore, the system can be simplified and a great economical effect can be obtained. Although the present invention has been described hereinbefore in the preferred forms, it is understood that the present disclosure of the preferred forms may be changed or modified in the details of construction, and the combination and arrangement thereof may be resorted to without departing from the spirit and the scope of the appended claims.
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	description	The United States Government has rights in this invention pursuant to Contract No. DE-AC07-05ID14517 between the U.S. Department of Energy (DOE) and Battelle Energy Alliance. The present invention relates to the separation and recovery of uranium from aluminum-clad metallic nuclear fuel. In an effort to reduce the use of highly enriched uranium fuels, different programs exist to convert research reactors fueled with highly enriched uranium into those fueled by low enriched uranium. The United States Highly Enriched Uranium Reactor Conversion program used to be known as the Reduced Enrichment for Research and Test Reactors program, and it is still known as such internationally. There remains a strong emphasis in converting research reactors fueled with highly enriched uranium into those fueled by low enriched uranium. The highly enriched uranium fuel is made of aluminide, oxide, or silicide particulate dispersed in an aluminum powder and clad with aluminum. The uranium loading densities range from 2 to 5 gU/cm3. Conversion to a low enriched uranium fuel, where there is less than 20% U-235, without impacting reactor performance requires a higher uranium loading density. 15.3 gU/cm3 is achievable within a monolithic U—Mo fuel. Current methods to recover uranium from high-enriched fuel use an aqueous process. During this process, the presence of uranium in combination with zirconium in monolithic uranium fuel introduces explosion hazards. The projected demand for monolithic low enriched uranium fuel within the U.S. is approximately 1,800 kg annually. Accounting for uranium losses during fabrication indicates an additional 3,000 kg monolithic low enriched uranium fuel would be required annually to meet projected demands. The used fuel and fabrication scrap are of sufficient U-235 enrichment to warrant recovery and reuse of the low enriched-uranium. Due to the high demand of low enriched uranium needed and the safety concerns associated with the aqueous process, there exists a need for a safe processing method of monolithic fuel that can meet the demand that is needed. According to one aspect of the invention, a method for separating and recovering uranium from aluminum-clad metallic nuclear fuel is described. The method includes immersing a nuclear fuel element containing nuclear fuel and cladding in a molten metal. The nuclear fuel includes uranium. The cladding is selectively dissolved from the nuclear fuel element when immersed in the molten metal. The nuclear fuel is separated from the cladding. The method then includes loading the nuclear fuel into a permeable basket that is electrically configured as an anode of an electrolytic cell. There are also a molten salt electrolyte and a cathode in the electrolytic cell. Then, the method includes applying an electric charge across the electrolytic cell. The molten salt electrolyte selectively transfers uranium from the anode to the cathode. According to another aspect of the invention, a method for separating uranium from a nuclear fuel element is described. The method includes immersing a nuclear fuel element containing nuclear fuel, an interlayer, and cladding in molten magnesium. The nuclear fuel is a uranium-molybdenum alloy fuel. The interlayer includes zirconium. The cladding includes aluminum. The cladding is selectively dissolved from the nuclear fuel rod when immersed in the molten magnesium. The nuclear fuel and interlayer are separated from the cladding. The method then includes loading the nuclear fuel and interlayer into a permeable basket that is electrically configured as an anode of an electrolytic cell. The electrolytic cell has a LiCl—KCl—UCl3 electrolyte and a cathode. Then, the method includes applying an electric charge across the electrolytic cell. The electrolyte selectively transfers uranium from the anode to the cathode. The following detailed description provides illustrations for embodiments of the present invention. Each example is provided by way of explanation of the present invention, not limitation of the present invention. Those skilled in the art will recognize that other embodiments for carrying out or practicing the present invention are also possible. Therefore, it is intended that the present invention covers such modifications and variations as come within the scope of the appended claims and their equivalents. Referring to FIG. 1, a flowchart illustrating a method 100 of separating and recovering uranium from a nuclear fuel rod is shown. A nuclear fuel element is immersed 102 into a molten metal. As shown in FIG. 2, the nuclear fuel element 200 contains nuclear fuel 202 and cladding 204. The nuclear fuel element 200 can be any nuclear fuel assembly, not limited to a specific shape. The fuel element 200 may be cylindrical or it may be an orthotope. The nuclear fuel 202 includes uranium and the nuclear fuel 202 may be monolithic. The nuclear fuel 202 may be irradiated or unirradiated. The nuclear fuel 202 may be a highly enriched uranium fuel. The nuclear fuel 202 may include fission products. In an embodiment, the nuclear fuel 202 is a monolithic uranium-molybdenum alloy fuel. In an embodiment, the cladding 204 contains aluminum or an aluminum alloy. FIG. 3 shows an embodiment of the nuclear fuel element 200 having an interlayer 206 between the nuclear fuel 202 and the cladding 204. In an embodiment, the interlayer 206 is a zirconium interlayer. Returning to FIG. 1, the molten metal is a liquid metal that can maintain the nuclear fuel in a metallic state while selectively dissolving the cladding. The molten metal targets the cladding for dissolution. In any embodiment, the cladding is soluble in the molten metal, but the nuclear fuel is not soluble in the molten metal. The molten metal has a melting point approximately below or near that of the cladding. In an embodiment where the cladding contains aluminum, the temperature of the molten metal is approximately 700° C. In an embodiment where the molten metal is molten magnesium, the melting point of magnesium is 650° C. and produces a molten Al—Mg alloy below 700° C. Further, magnesium does not interact with uranium, molybdenum, or zirconium; meaning the nuclear fuel, and associated interlayer if present, would be undisturbed. In an embodiment where the molten metal is molten lithium, the melting point of lithium is 180° C. and produces a molten Al—Li alloy below 700° C. Similarly, lithium does not interact with uranium, molybdenum, or zirconium; meaning the nuclear fuel, and associated interlayer if present, would be undisturbed. When the nuclear fuel element is immersed 102 in the molten metal, the cladding is dissolved from the nuclear fuel element. When dissolved, the cladding forms a molten alloy with the molten metal and the nuclear fuel is separated from the cladding, leaving the nuclear fuel substantially free of cladding. Substantially free of cladding means that most, at least approximately 60%, of the cladding is removed from the nuclear fuel. Preferably, approximately 90% of the cladding is removed from the nuclear fuel. In an embodiment, the nuclear fuel is dried after immersing the nuclear fuel rod into the molten metal. The nuclear fuel would be dried through a vacuum distillation or any method that provides acceptable results. In prior art methods of using an aqueous solution to remove cladding, explosion hazards existed, especially if there was a zirconium interlayer. In prior art methods, the outcomes were poor due to oxidation of the metal fuel matrix in an aqueous solution, which inhibited the subsequent recovery of uranium metal. The present invention does not pose an explosion hazard because the molten metal does not interact with the nuclear fuel or the interlayer. The present invention causes no oxidation of the nuclear fuel. Unlike prior art methods, where a caustic dissolution is used to remove an aluminum cladding, the present invention is safer and more effective. In an experiment, five strips of aluminum clad U—Mo alloy fuel were used. The five strips were approximately 2 in long, ¼ in wide, and 0.6 in thick. The strips were immersed in a molten magnesium at 700° C. FIG. 4 shows the phase diagram for aluminum in magnesium. As shown in the phase diagram, aluminum is soluble in magnesium. FIG. 5, FIG. 6, and FIG. 7 show the phase diagrams for magnesium in zirconium, uranium and molybdenum, respectfully. As these diagrams depict, and as the experiment confirmed, magnesium is not soluble in zirconium, uranium and molybdenum. In application, this resulted in the aluminum cladding dissolving within a matter of seconds after contacting the molten magnesium. The U—Mo alloy fuel strips were left intact and undisturbed by the removal process of the cladding. The nuclear fuel is loaded 104 into a permeable basket. The permeable basket is electrically configured as an anode within an electrolytic cell. The anode is the electrode through which an electric current flows within a polarized electrolytic cell. The anode is where the chemical oxidation occurs. The anode is made from an electrically conductive or semiconductive material. In an embodiment the anode is a metal basket. For example, the anode may be made from stainless steel. The electrolytic cell also has a molten salt electrolyte and a cathode. The molten salt electrolyte is any liquid salt electrolyte that is ion selective towards uranium. The molten salt electrolyte could include various combinations of molten halides, as long as the companion uranium trihalide is present within the molten salt electrolyte. In some embodiments, the molten salt electrolyte is a combination of the same halide. For example, the halide can be chloride. Or the halide can include alkali chlorides (i.e., LiCl, KCl, NaCl). The molten salt electrolyte can be LiCl—KCl—UCl3. The cathode is the electrode through which an electric current flows to within a polarized electrolytic cell. The cathode is where the chemical reduction occurs. The cathode is made from an electrically conductive or semiconductive material. In an embodiment the cathode is a metal rod. For example, the cathode may be made from stainless steel. When an electric charge is applied 106 across the cell, the molten salt electrolyte selectively transfers uranium from the anode to the cathode. The electric charge can be a controlled current or controlled electric potential. The molten salt electrolyte oxidizes the uranium ions at the anode when the electric charge is applied. The molten salt electrolyte will selectively filter out contaminates by only transferring uranium to the cathode. The molten salt electrolyte reduces the uranium ions at the cathode to uranium metal. The cathode deposit of uranium metal is separated from the molten salt electrolyte. Separation can be by lifting the cathode from the molten salt electrolyte or by removing the molten salt electrolyte from the cell. The separated uranium is devoid of other fuel constituents. Other fuel constituents include fission products, alloying agents, remnant cladding pieces, or other non-nuclear fuel parts of the nuclear fuel element. In an embodiment, the uranium at the cathode is recovered and used for fabrication in a low enriched uranium fuel. Low enriched uranium fuel is uranium fuel with less than approximately a 20% enrichment of U-235. It is to be understood that the above-described arrangements are only illustrative of the application of the principles of the present invention. Numerous modifications and alternative arrangements may be devised by those skilled in the art without departing from the spirit and scope of the present invention and the appended claims are intended to cover such modifications and arrangements. Any element in a claim that does not explicitly state “means for” performing a specified function, or “step for” performing a specific function, is not to be interpreted as a “means” or “step” clause as specified in 35 U.S.C. § 112, ¶6. In particular, the use of “step of” in the claims herein is not intended to invoke the provisions of 35 U.S.C. § 112, ¶6.
	description	This application claims the benefit of U.S. Provisional Application No. 61/020,054, filed Jan. 9, 2008. Not Applicable 1. Field of Invention This invention relates to a coil diagnostic system for monitoring rod position indicator coils in a nuclear power plant. More specifically, this invention relates to a system performing on-line rod coil diagnostics for rod position indication systems. 2. Description of the Related Art In a Pressurized Water Reactor (PWR), the power level of the reactor 10 is controlled by inserting and retracting the control rods 12, which for purposes of this application include the shutdown rods, into the reactor core 14. The control rods are moved by the Control Rod Drive Mechanisms (CRDM), which are electromechanical jacks that raise or lower the control rods in increments. The CRDM includes a lift coil DML, a moveable gripper coil DMM, and a stationary gripper coil DMS that are controlled by the Rod Control System (RCS) and a ferromagnetic drive rod that is coupled to the control rod and moves within the pressure housing 16. The drive rod includes a number of circumferential grooves at ⅝ inch intervals (“steps”) that define the range of movement for the control rod. A typical drive rod contains approximately 231 grooves, although this number may vary. The moveable gripper coil mechanically engages the grooves of the drive rod when energized and disengages from the drive rod when de-energized. Energizing the lift coil raises the moveable gripper coil (and the control rod if the moveable gripper coil is energized) by one step. Energizing the moveable gripper coil and de-energizing the lift coil moves the control rod down one step. Similarly, when energized, the stationary gripper coil engages the drive rod to maintain the position of the control rod and, when de-energized, disengages from the drive rod to allow the control rod to move. The RCS includes the logic cabinet and the power cabinet. The logic cabinet receives manual demand signals from an operator or automatic demand signals from Reactor Control and provides the command signals needed to operate the shutdown and control rods according to a predetermined schedule. The power cabinet provides the programmed dc current to the operating coils of the CRDM. Current PWR designs have no direct indication of the actual position of each control rod. Instead, step counters associated with the control rods are maintained by the RCS and rod position indication (RPI) systems to monitor the positions of the control rods within the reactor. The associated step counter is incremented or decremented when movement of a control rod is demanded and successful movement is verified. Because the step counter only reports the expected position of the control rod, certain conditions can result in the step counter failing and deviating from the actual position of the control rod. In certain situations where the actual position of the control rod is known, the step counter can be manually adjusted to reflect the actual position. However, if the actual position of the control rod is not known, a plant shutdown may be required so that the step counters to be initialized to zero while the control rods are at core bottom. The RPI systems derive the axial positions of the control rods by direct measurement of drive rod positions. Currently both analog rod position indication (ARPI) systems and digital rod position indication (DRPI) systems are in use in PWRs. The conventional DRPI systems have been in service for over 30 years in nuclear power stations worldwide and are currently being used as the basis for the rod position indication systems in the new Westinghouse AP1000 designs. A conventional DRPI system includes two coil stacks for each control rod and the associated DPRI electronics for processing the signals from the coil stacks. Each coil stack is an independent channel of coils placed over the pressure housing. Each channel includes 21 coils. The coils are interleaved and positioned at 3.75 inch intervals (6 steps). The DRPI electronics for each coil stack of each control rod are located in a pair of redundant data cabinets (Data Cabinets A and B). Although intended to provide independent verification of the control rod position, conventional RPI systems are not accurate to fewer than 6 steps. The overall accuracy of a DRPI system is considered to be accurate within ±3.75 inches (6 steps) with both channels functioning and ±7.5 inches using a single channel (12 steps). In contrast to the conventional DRPI system, a conventional ARPI system determines the position based on the amplitude of the dc output voltage of an electrical coil stack linear variable differential transformer. The overall accuracy of a properly calibrated ARPI system is considered to be accurate within ±7.2 inches (12 steps). Neither conventional ARPI systems nor conventional DRPI systems are capable of determining the actual positions of the control rods. In the event of a step counter failure, plant shutdown for re-initialization of the step counter is still required because the approximate positions of the control rods reported by conventional RPI are of little or no value. It should be noted that for purposes of this application, the phrase “control rod” is used generically to refer to a unit for which separate axial position information is maintained, such as a group of control rods physically connected in a cluster assembly. The number of control rods varies according to the plant design. For example, a typical four-loop PWR has 53 control rods. Each control rod requires its own sets of coils having one or more channels and the DRPI electronics associated with each channel. Thus, in a typical four-loop PWR, the entire DPRI system would include 53 coil stacks, each having two independent channels, and 106 DPRI electronics units. Further, in this application, the phrase “coil stack” is used generically to refer to the detector coils associated with each control rod and should be understood to include either or both channels of detector coils. Thus, a measurement across a coil stack contemplates the value across both channels combined and/or the value across a single channel. Unfortunately, aging and obsolescence issues have led to an increase in problems with conventional DRPI systems including analog card failures and coil cable connection problems that, in some cases, may result in unplanned reactor trips. These problems, along with plans for plant life extension, have prompted the industry to actively seek viable options to monitor the health and accuracy of the DRPI systems and/or to replace failing systems in order to ensure reliable plant operations for decades to come. In addition to obsolescence concerns, the lack of diagnostic capabilities is a significant problem. Conventional RPI systems cannot provide any diagnostic information on their health other than the current rod position indication. Accordingly, diagnostics of the RPI system is limited to periods when the PWR is offline. Currently, the offline RPI coil diagnostic procedures include performing resistance and inductance measurements at high frequencies (frequencies above line frequency, e.g., 100, 1,000, and 10,000 hertz) on the RPI coils during each refueling outage after the reactor head is re-connected and before the RPI system is energized using traditional inductance, capacitance, and resistance (LCR) meters. By way of example, in a four loop PWR with 53 control rod assemblies, performing diagnostics on each coil requires a total of 2226 measurements. This traditional testing takes 8-12 hours and costs hundreds of thousands of dollars. Moreover, this testing may be skipped during some outages in an effort to reduce operating costs and/or plant downtime. The primary benefit of offline diagnostics is to catch obvious failures resulting from reassembly of the reactor. However, in between refueling outages, RPI failures can occur without warning, which leads to increased costs for the plant, especially if replacement parts cannot be obtained in a timely manner. Without active monitoring, plant engineers cannot identify problems developing in the RPI systems and take preemptive actions, such as obtaining necessary replacement parts ahead of time and replacing failing components at the next scheduled outage; rather plants must begin remedial actions after an actual failure occurs. Beyond the technical problems of the conventional DRPI systems, regulatory issues exist. Many existing PWRs are approaching the end of qualified life for several components of the conventional DRPI systems during the next decade and are actively seeking replacement options at this time. There has been a significant push in recent years for plants to replace aging analog systems with digital systems made from commercially-available off-the-shelf parts. Using readily-available commercial parts provide plants more options for replacement in the future. An automated system for on-line monitoring and coil diagnostics of rod position indicator (RPI) coils coil diagnostic, or RPI coil diagnostic system, is described herein. The RPI coil diagnostic system performs coil diagnostics for a RPI system in a nuclear power plant. The RPI coil diagnostic system is in electrical communication with and monitors the outputs of the detector coils. The RPI coil diagnostic system measures characteristics of the detector coils that are indicative of the health of the detector coils and/or the connections between the detector coils and the RPI electronics. The RPI coil diagnostic system can be implemented as a complete system in a new plant design or a supplemental system that works in conjunction with portions of a conventional DRPI system to provide position measurements with improved resolution compared to the conventional DRPI. The RPI coil diagnostic system includes RPI coil diagnostic electronics that are connected to and monitor the electrical signals from the plurality of detector coils and the reference voltage. The RPI coil diagnostic electronics include a data acquisition unit in communication with an interface device. The RPI coil diagnostic data acquisition unit has a number of analog inputs equal to the number of coils in a single channel plus an additional input for the reference line. The electronic signals produced by each DRPI coil are sampled by RPI coil diagnostic data acquisition unit. The RPI coil diagnostic processing unit receives RPI coil diagnostic data from the RPI coil diagnostic electronics including ac voltage measurements and ac current measurements. The RPI coil diagnostic processing unit uses the RPI coil diagnostic data to calculate the impedance at each of the test points by dividing the complex voltage measurements by the complex current measurements. Because, the excitation frequency of the coils is known (to be the line frequency), the RPI coil diagnostic processing unit derives the respective resistances and reactive inductances for the detector coils from the calculated impedance. Deviations in the impedances, resistances, inductances from baseline values or the expected linear relationships within a coil set indicate a potential or actual problem with the detector coil or its associated connections. An automated system for on-line monitoring and coil diagnostics of rod position indicator (RPI) coils coil diagnostic, or RPI coil diagnostic system, is described in detail herein and illustrated in the accompanying figures. The RPI coil diagnostic system performs diagnostics for a digital rod position indication (DRPI) system in a nuclear power plant. The RPI coil diagnostic system is in electrical communication with and monitors the outputs of the detector coils. The RPI coil diagnostic system measures identifiable electrical characteristics of the detector coils that are indicative of the health of the detector coils and/or the connections between the detector coils and the DRPI system. The identifiable electrical characteristic can be inductance, impedance, resistance, or other electrical characteristics which are measurable for diagnostic purposes. FIG. 1 is a block diagram of an RPI coil diagnostic system in a pressurized water reactor (PWR). A brief overview of the systems of a PWR that are relevant to the RPI coil diagnostic system can be found in the description of the related art. The RPI coil diagnostic system can be implemented as a complete system in a new plant design or a supplemental system that works in conjunction with portions of a conventional RPI system to provide self-diagnostic capabilities not available with a conventional RPI system. In the illustrated embodiment, the RPI coil diagnostic system includes the RPI coil diagnostic electronics located inside containment and the RPI coil diagnostic processing unit located outside containment in the main control room. The RPI coil diagnostic electronics sample the electrical signals from the detector coils and transmit the sampled data to the RPI coil diagnostic processing unit. The RPI coil diagnostic processing unit evaluates the sampled data from the RPI coil diagnostic electronics to determine the health of the detector coils and/or the associated connections. The diagnostic information generated by the RPI coil diagnostic processing unit is displayed to the reactor operators via a user interface and may be used to identify deteriorating components and other problems prior to actual failure. FIG. 2 illustrates one embodiment of the RPI coil diagnostic system used to retrofit plants with existing conventional DRPI systems. The conventional DRPI system consists of two redundant components (Data Cabinets A and B) located inside the containment area and in communication with the detector coils of the coil stacks mounted on the rod control housing above the reactor. In this embodiment, the RPI coil diagnostic electronics are connected to the data cabinets at a point between the input from the existing detector coils and the conventional DRPI electronics allowing the RPI coil diagnostic electronics to sample the DRPI coil currents and convert them into digital signals. The digital signal is then transmitted to the RPI coil diagnostic processing unit in the main control room. FIG. 3 is a diagram of one embodiment of the RPI coil diagnostic electronics used in the retrofit application of FIG. 2. In this embodiment, the RPI coil diagnostic electronics are connected to the test points PT1-PTn, PTREF in the data cabinets of the conventional DRPI. The test points PT1-PTn, PTREF provide access to the electrical signals from the plurality of detector coils C1-Cn and the reference voltage VREF. In the prior art, the primary use of the test points PT1-PTn, PTREF is for manual diagnostics of the detector coils C1-Cn and the connections PC1-PCn, PCREF when the nuclear power plant is offline. As previously discussed, these manual offline diagnostics occur at higher frequencies as compared to the operating frequency (i.e., 60 Hz in the U.S.) and require the technicians to connect 2,226 different sets of test points to the LCR meter. The RPI coil diagnostic electronics include a data acquisition unit in communication with an interface unit. Each control rod has one RPI coil diagnostic electronics unit for each independent channel of the coil stack associated with the control rod. For example, a PWR having 53 control rods monitored by redundant DPRI systems (53 coil stacks with two independent channels) would have 106 ADPRI electronics (53 per data cabinet). In one embodiment, each ADRPI data acquisition unit has a number of analog inputs equal to the number of coils in a single channel plus an additional input for the reference line. The electronic signals produced by each DRPI coil are sampled by RPI coil diagnostic data acquisition unit. The interface unit is used to transmit the sampled data to the RPI coil diagnostic processing unit located outside containment. The interface unit is selected to have sufficient data transmission speeds to send the sampled data to the RPI coil diagnostic processing unit in real time. By way of example, one suitable device for performing the functions of the RPI coil diagnostic data acquisition unit and the interface unit is the CompactRIO remote high speed interface system produced by National Instruments Corporation, which includes swappable I/O modules connected to an FPGA for acquiring various types of signals including the voltage and current signals used by the RPI coil diagnostic system and a high speed interface allowing an external computer to communicate with the FPGA at data rates up to 50 MB/s. One skilled in the art will recognize that the general specifications for the RPI coil diagnostic electronics are not intended to be limiting and that deviations intended to acquire sufficient data containing information from which the positions of the control rods to a single step can be derived are considered to remain with the scope and spirit of the appended claims. The RPI coil diagnostic data acquisition unit and the interface receive the electrical signals from the plurality of DRPI coils when the DRPI coils are energized. In other words, the RPI coil diagnostic system begins operating during plant startup when the DRPI system is energized and continues during normal operation of the PWR. Accordingly, the RPI coil diagnostic system begins sampling the electrical signals received from all of the detector coils early in the startup phase. Within a span of a few seconds to a few minutes, the RPI coil diagnostic system can process the sampled data from all detector coils and identify actual or potential problems with the approximately 2,226 detector coils and/or the connections thereto. Once the PWR is online, the RPI coil diagnostic system continues diagnostic monitoring as it continues to sample and process the electrical signals from the detector coils allowing actual or potential problems with the detector coils to be identified before a failure scenario occurs. As a result, preventative maintenance can occur during regularly scheduled outages rather than during an unplanned shutdown from a reactor trip to a failure in the DRPI system. Under normal plant operating conditions, the diagnostic measurements are made at the operating frequency supplied by the detector coil power supply, which will typically be a low voltage at either 50 or 60 hertz depending upon the locality of the PWR. At times when the DRPI system is not energized, for example, during a plant shutdown, the RPI coil diagnostic system can be used to rapidly perform diagnostics on some or all of the detector coils by energizing the selected coils. In one embodiment, the detector coils are energized by an external power supply. Use of an external power supply also allows the selection of a different frequency voltage source. FIG. 4 illustrates one embodiment of the process applied by the processing unit of the RPI coil diagnostics system. The RPI coil diagnostic processing unit receives RPI coil diagnostic data from the RPI coil diagnostic electronics including ac voltage measurements. More specifically, the ac voltages at the test points PT1-PTn, PTREF are measured. The ac currents at the test points PT1-PTn may be measured or calculated. In one embodiment, the ac current at the test points PT1-PTn are calculated from the ac voltages because the resistance from the test points PT1-PTn to ground is known. The ac current is calculated by: I ~ Tn = V ~ Tn R Tn . ( 1 ) In another embodiment, the RPI coil diagnostic system measures the coil currents directly eliminating the need to calculate them. Using the RPI coil diagnostic data, the RPI coil diagnostic processing unit calculates the impedances at each of the test points PT1-PTn associated with the detector coils. The RPI coil diagnostic processing unit calculates the impedance according to the equation: Z ~ n = ( V ~ REF - V ~ Tn ) I ~ Tn . ( 2 ) The RPI coil diagnostic processing unit uses the calculated impedances to derive the respective resistances and reactive inductance for the detector coils C1-Cn. In one embodiment where the detectors coils are energized by a known source, such an ac voltage source at line frequency, the RPI coil diagnostic processing unit calculates the resistances and inductances using the known line frequency. In another embodiment where the source frequency is unknown, the RPI coil diagnostic processing unit determines the frequency at the reference point PTREF. The RPI coil diagnostic processing unit calculates the resistance and inductance for each the detector coils C1-Cn by the equation:{tilde over (Z)}n=Rn+j(2πfvREFLn). (3)A high resistance indicates a bad connection. A low resistance or inductance indicates shorted turns in the detector coil winding. To identify problems, either actual or potential, the calculated resistance and inductance are compared to a reference resistance and reference inductance for each of the detector coils C1-Cn. In one embodiment, the RPI coil diagnostic system is calibrated by moving the control rods through their entire range of motion and obtaining and storing baseline values for the impedances. Any or all of the calculated or baseline/reference resistances, inductances, and the impedances for the detector coils C1-Cn are displayed for the plant technicians at the user interface. The calibration process may be repeated and the results averaged, if necessary, to create an accurate set of baseline impedances. Thus, the baseline may include both the impedance when the drive rod passes through the detector coil and the impedance when the drive rod is not in the detector coil. In another embodiment, problems are detected without the benefit of baseline data simply by looking for deviations in the relationships of the sampled data. A prototype of the RPI coil diagnostic system was tested at the Farley nuclear power plant using a single channel of detector coils for one control rod when withdrawing the control rod 226 steps out of core, inserting the control rod 226 steps into the core, and during rod drop testing. Tables 1 and 2 are based on the data obtained during testing of the prototype. Table 1 shows the impedance, the resistance, and the inductance at a frequency of 60 hertz, calculated as described above while the control rod is fully inserted into the reactor core. Within a coil stack, the resistance values, the inductance values, and the impedance values have a substantially linear relationship. During the operation of a DRPI system, the detector coil impedances (and resistances and inductances) vary based on the positions of the control rods. In Table 1, the values for detector coil 1 deviate from the substantially linear relationship of detector coils 2-21 because the drive rod continues to influence the magnetic field of detector coil 1 when the control rod is fully inserted into the reactor core (i.e., the rod remains within Coil 1 when fully inserted). TABLE 1Data Set #1: Resistance, Inductance, and Impedance @ 60 hertzDetectorResistanceInductanceImpedanceCoil(Ohms)(mH)(Ohms)113.36 36.0813.36 + j13.60 28.1542.678.15 + j16.0937.4042.097.40 + j15.8747.6142.267.61 + j15.9357.4442.467.44 + j16.0167.5142.227.51 + j15.9277.6842.207.68 + j15.9187.5741.837.57 + j15.7797.5442.207.54 + j15.91107.7442.167.74 + j15.89117.7741.747.77 + j15.74127.7141.647.71 + j15.70137.7541.447.75 + j15.62147.7041.397.70 + j15.60157.9541.397.95 + j15.61167.8241.077.82 + j15.48178.0240.568.02 + j15.29187.8540.097.85 + j15.11198.0439.678.04 + j14.96208.0239.138.02 + j14.75217.9638.257.96 + j14.42 Table 2 shows data for a coil stack with the control rod fully withdrawn from the reactor core. Deviations from the linear relationship are indicative of a potential or actual problem. From the data, it can be seen that the resistance of coil 4 is anomalous, indicating a potential problem with that coil or the associated connections. In this instance, the deviation is not sufficient to impair the DRPI operation or to cause a reactor trip but warrants further monitoring. TABLE 2Data Set #2: Resistance, Inductance, and Impedance @ 60 hertzDetectorResistanceInductanceImpedanceCoil(Ohms)(mH)(Ohms)16.3942.556.39 + j16.0426.4944.196.49 + j16.6636.5844.176.58 + j16.6548.7645.898.76 + j17.3056.6444.396.64 + j16.7366.6343.996.63 + j16.5876.7244.096.72 + j16.6286.6943.956.69 + j16.5796.8044.216.80 + j16.67106.8543.466.85 + j16.38116.8543.716.85 + j16.48126.9743.566.97 + j16.42136.9643.156.96 + j16.27147.1443.797.14 + j16.51157.0843.207.08 + j16.29167.0342.427.03 + j15.99177.0742.307.07 + j15.95187.1042.187.10 + j15.90197.1042.547.10 + j16.04207.1542.027.15 + j15.84218.1639.068.16 + j14.73 In alternate embodiments, the RPI coil diagnostic data may include measurements at other frequencies or at other points which provide for calculation of other electrical characteristics which are measurable for diagnostic purposes without departing from the spirit and scope of the present invention. Measurements obtained while exciting the detector coils with a dc source would be the equivalent of an ac source with a frequency of zero hertz and would produce an impedance without any significant imaginary component. In other words, the impedance would simply appear as the dc resistance. FIG. 5 illustrates an alternate embodiment of the RPI coil diagnostic system that completely replaces conventional DRPI systems. In this embodiment, the RPI coil diagnostic electronics are connected directly to the DPRI coils and communicate directly with the RPI coil diagnostic processing unit located outside containment. The RPI coil diagnostic system may also replace a conventional ARPI system; however, as previously discussed, the detector of a conventional ARPI system differs from the DRPI coils used for drive rod presence sensing. Accordingly, when replacing a conventional ARPI system, the RPI coil diagnostic system necessarily includes the DRPI detection coils. However, in existing ARPI systems the RPI coil diagnostic system can be used, solely, for obtaining measurements from the detector coils for performing coil diagnostics. FIG. 6 illustrates an alternate embodiment of the RPI coil diagnostic system where some or all of the processing functions occur in the RPI coil diagnostic electronics. In the embodiment of FIG. 6, the RPI coil diagnostic electronics includes a processing unit receiving data from the RPI coil diagnostic data acquisition unit. In one embodiment, the processing unit in the RPI coil diagnostic electronics calculates the impedance thereby reducing the amount of data that must be transferred to the main control system. In another embodiment, the processing unit in the RPI coil diagnostic electronics assumes all of the processing functions thereby eliminating the need for the RPI coil diagnostic system to provide a separate processing unit in the main control system. In this embodiment, the RPI coil diagnostic electronics communicate directly with other control systems in the PWR, such as the reactor control of the main control system or the logic cabinet in the rod drive system, or simply communicates with the user interface. For example, in a system using the CompactRIO previously described, the FPGA calculates the impedance. One skilled in the art will appreciate that the processing units described herein can be implemented using any number of logic components including controllers and processors without departing from the scope and spirit of the present invention. From the foregoing description, it will be recognized by those skilled in the art that an RPI coil diagnostic system capable of monitoring the health of the detector coils and the associated connections in a PWR has been provided. The RPI coil diagnostic system is capable of providing diagnostic information about the health of the detector coils and the associated connections while the plant is both online and offline. The diagnostic information includes both historical/baseline values for and real-time determination of the impedance, resistance, and/or inductances of the detector coils. The RPI coil diagnostic system makes the diagnostic information available to the plant operators on a real-time basis. The RPI coil diagnostic system is capable of generating alerts based on the diagnostic information when deviations from the baseline values are detected. While the present invention has been illustrated by description of several embodiments and while the illustrative embodiments have been described in considerable detail, it is not the intention of the applicant to restrict or in any way limit the scope of the appended claims to such detail. Additional advantages and modifications will readily appear to those skilled in the art. The invention in its broader aspects is therefore not limited to the specific details, representative apparatus and methods, and illustrative examples shown and described. Accordingly, departures may be made from such details without departing from the spirit or scope of applicant's general inventive concept.
	claims	1. A collimator assembly comprising:a plurality of blades arranged in parallel with respect to one another;wherein each of the plurality of blades comprises a body and a respective plurality of fins coupled to the blade and extending therefrom;wherein each of the respective plurality of fins is coupled to an adjacent one of the plurality of blades; andwherein the body of each of the plurality of blades comprises a different material than the respective plurality of fins. 2. The collimator assembly, as set forth in claim 1, wherein each of the respective plurality of fins is perpendicular to the body of the blade and parallel to each other. 3. The collimator assembly, as set forth in claim 1, wherein at least a first number of each of the respective plurality of fins are arranged perpendicular to the body of the blade, and at least a second number of the plurality of blades extends at an angle of between 0 and 10 degrees from perpendicular with respect to the body. 4. The collimator assembly, as set forth in claim 1, wherein each of the respective plurality of fins extends from only one side of the blade. 5. The collimator assembly, as set forth in claim 1, wherein a first number of each of the respective plurality of fins extends from a first side of the blade and a second number of each of the respective plurality of finds extends from a second side of the blade. 6. The collimator assembly, as set forth in claim 1, wherein each of the respective plurality of fins is coupled to the respective plurality of fins on an adjacent one of the plurality of blades. 7. The collimator assembly, as set forth in claim 1, wherein each of the respective plurality of fins extends a length in the range of approximately 0.4 mm to about 1.5 mm from the blade. 8. The collimator assembly, as set forth in claim 1, wherein each of the plurality of fins comprises a polymer material loaded with a high atomic number material. 9. A collimator assembly comprising:a first blade having a first plurality of fins extending therefrom;a second blade arranged in parallel to the first blade and having a second plurality of fins extending therefrom; andwherein a body of each of the first and second blades comprises a different material than the respective first and second plurality of fins. 10. The collimator assembly, as set forth in claim 9, wherein the first plurality of fins is perpendicular to a body of the first blade and parallel to each other and wherein the second plurality of fins is perpendicular to a body of the second blade and parallel to each other. 11. The collimator assembly, as set forth in claim 9, wherein at least a number of the first plurality of fins extends from a body of the first blade at an angle of between 0 and 10 degrees from perpendicular with respect to the body of the first blade, and wherein at least a number of the second plurality of fins extends from a body of the second blade at an angle of between 0 and 10 degrees from perpendicular with respect to the body of the first blade. 12. The collimator assembly, as set forth in claim 9, wherein each or the first plurality or fins extends from only one side of the first blade, and wherein each of the second plurality of fins extends from only one side of the second blade. 13. The collimator assembly, as set forth in claim 9, wherein a first number of the first plurality of fins extends from a first side of the first blade and wherein a second number of the first plurality of fins extends from a second side of the first blade, and wherein a first number of the second plurality of fins extends from a first side of the second blade and wherein a second number of the second plurality of fins extends from a second side of the second blade. 14. The collimator assembly, as set forth in claim 9, wherein the first plurality of fins is coupled to the second plurality of fins. 15. The collimator assembly, as set forth in claim 9, wherein the second plurality of fins is coupled to the first blade. 16. An imaging detector assembly comprising:a detector array;a scintillator assembly configured to receive incident radiation and configured to convert the incident radiation to light for transmission to the detector array; anda collimator assembly comprising:a plurality of blades arranged in parallel with respect to one another;wherein each of the plurality of blades comprises a respective plurality of fins coupled to the blade and extending there from;wherein each of the respective plurality of fins is coupled to an adjacent one of the plurality of blades; andwherein a body of each of the plurality of blades comprises a different material than the respective plurality of fins. 17. The imaging detector assembly, as set forth in claim 15, wherein each of the respective plurality of fins extends at an angle of approximately 90° with respect to the blade. 18. The imaging detector assembly, as set forth in claim 15, wherein at least a number of each of the respective plurality of fins are arranged at angles between approximately 85° and 95° relative to the length of the blade. 19. The imaging detector assembly, as set forth in claim 15, wherein each of the respective plurality of fins extends from only one side of the blade. 20. The imaging detector assembly, as set forth in claim 15, wherein a first number of each of the respective plurality of fins extends from a first side of the blade and a second number of each of the respective plurality of finds extends from a second side of the blade. 21. The imaging detector assembly, as set forth in claim 15, wherein each of the respective plurality of fins is coupled to the respective plurality of fins on an adjacent one of the plurality of blades. 22. The imaging detector assembly, as set forth in claim 15, wherein each of the respective plurality of fins extends a length in the range of approximately 0.4 mm to about 1.5 mm from the blade. 23. A method of fabricating a collimator assembly comprising:forming a plurality of blades, wherein each of the plurality of blades comprises a respective plurality of fins extending there from;coupling each of the respective plurality of fins to an adjacent one of the plurality of blade; andwherein a body of each of the plurality of blades comprises a different material than the respective plurality of fins. 24. The method, as set forth in claim 23, wherein forming each of the plurality of blades comprises forming a blade having fins extending from only one side of the blade. 25. The method, as set forth in claim 23, wherein forming each of the plurality of blades comprises forming a blade having fins extending from each side of the blade. 26. The method, as set forth in claim 23, wherein forming each of the plurality of blades comprises coupling the respective plurality of fins to the blade. 27. The method, as set forth in claim 23, wherein coupling each of the respective plurality of fins to an adjacent one of the plurality of blades comprises coupling each of the respective fins of the associated one of the plurality of blades to the respective fins of an adjacent one of the respective plurality of blades.
	abstract	A passive containment air cooling system for a nuclear power plant that enhances air flow over a metal containment that houses the reactor system to improve heat transfer out of the containment. The heat transfer is improved by employing swirl vanes to mix the air as it rises over the walls of the containment due to natural circulation and a vortex engine proximate an exit along the cooling air path to increase the quantity of air drawn along the containment.
	claims	1. A process for introducing a reactor coolant additive in a pressurized water reactor having a primary circuit, a reactor core and reactor coolant water passing through the primary circuit, to produce elemental carbon, comprising:adding a sufficient amount of the reactor coolant additive to the reactor coolant water passing through the primary circuit of the pressurized water reactor, the reactor coolant additive being an organic compound selected from the group consisting of:elements of carbon and hydrogen,elements of carbon, hydrogen and nitrogen,elements of carbon, hydrogen and oxygen, andelements of carbon, hydrogen, nitrogen and oxygen,wherein said reactor coolant water excludes the presence of inorganic compounds with the exception of a sufficient amount of boric acid to control reactivity, hydrogen to provide reducing conditions, an additive to maintain pH in a target control band and trace elements naturally occurring in water. 2. The process of claim 1, wherein the equivalent elemental carbon addition rate is maintained in a range of from about 1 mg/hour to about 10 g/hour. 3. The process of claim 1, wherein the reactor coolant water is in a reactor coolant system of a nuclear reactor. 4. The process of claim 1, wherein the organic compound is selected from the group consisting of organic acids, alcohols, amines, aldehydes, ketones, and mixtures thereof. 5. The process of claim 1, wherein the organic compound is selected from the group consisting of acetic acid, methanol, ethanol, ethylamine, ethanolamine, and mixtures thereof. 6. The process of claim 1, wherein the organic compound is substantially soluble. 7. The process of claim 1, further comprising producing corrosion product deposits in the reactor core comprising elemental carbon in a range of from about 15 to about 20 percent by weight of the deposits. 8. The process of claim 1, wherein a radiation level in the reactor core is up to about 4000 Mrad/hour from gamma and neutrons. 9. The process of claim 1, further comprising maintaining a hydrogen concentration in the reactor core of greater than 0 cc/kg. 10. The process of claim 1, further comprising maintaining a hydrogen concentration in the reactor core of from about 25 to about 50 cc/kg. 11. The process of claim 1, wherein adding the organic compound is on a continuous basis. 12. The process of claim 1, wherein adding the organic compound is on a batch basis. 13. The process of claim 1, wherein the organic compound is in a high purity form consistent with standard nuclear industry practice of limiting impurities to as low as reasonably achievable (ALARA) for any additive to the coolant water of the pressurized water reactor. 14. The process of claim 1, further comprising producing corrosion product deposits in the reactor core wherein the elemental carbon is produced in an amount effective to change at least one of the morphology, deposition pattern, residence time, and carbon content of crud deposits in the reactor core as a result of adding the organic compound. 15. The process of claim 1, further comprising producing corrosion product deposits in the reactor core wherein the elemental carbon is produced in an amount effective to inhibit at least one of crud induced power shift, crud induced localized corrosion, cladding corrosion in the reactor core, and fuel failures as a result of adding the organic compound. 16. A process for introducing a coolant water additive to produce elemental carbon in a nuclear reactor having a primary circuit, comprising:adding a sufficient amount of the coolant water additive to coolant water passing through the primary circuit of the nuclear reactor, the coolant water additive being an organic compound selected from the group consisting of:elements of carbon and hydrogen,elements of carbon, hydrogen and nitrogen,elements of carbon, hydrogen and oxygen, andelements of carbon, hydrogen, nitrogen and oxygen,wherein said reactor coolant water excludes the presence of inorganic compounds with the exception of a sufficient amount of boric acid to control reactivity, hydrogen to provide reducing conditions, an additive to maintain pH in a target control band and trace elements naturally occurring in water. 17. The process of claim 16, wherein the equivalent elemental carbon addition rate is maintained in a range of from about 1 mg/hour to about 10 g/hour. 18. The process of claim 16, wherein the nuclear reactor is a pressurized water nuclear reactor.
	claims	1. A Thorium molten salt energy system comprising:a proton beam source for producing a proton beam having an energy level, wherein the proton source is adapted to vary the energy level of the produced proton beam between at least a first energy level and a second energy level, the proton beam source having a power input to receive power for driving the proton beam source,wherein the first energy level is greater than the second energy level, wherein the first energy level is such that a proton at the first energy level can interact with a Beryllium nucleus to produce a (p, n) reaction resulting in the generation of a neutron at an energy level sufficient to fission Thorium;wherein the second energy level is such that the interaction of a proton at the second energy level can interact with a Lithium nucleus to produce a (p, n) reaction resulting in the generation of a neutron at an energy level sufficient to fission Uranium;a Thorium molten salt assembly comprising:a main assembly body;a tubular member positioned within the main assembly body,a top lid coupled to the main assembly body in the form of a circular disk defining a plurality of openings passing therethrough; the plurality of openings including:a window opening passing through the top lid having a diameter larger than the other openings, the window opening defining a top window opening through which protons from the proton source may pass;a first group of openings comprising at least two impeller openings passing through the top lid each opening defining an opening suitable for receipt of a rotating shaft;a second group of openings comprising at least two heat exchanger openings passing through the top lid each opening configured for receipt of either an input or an output pipe element of a primary heat exchanger;a third group of openings comprising at least eight bolt openings passing through the lid each opening defining a bolt opening; anda window element positioned within the window opening; the window configured to permit the passage of protons therethrough;a molten salt solution contained within the main assembly body, the molten salt solution containing Thorium and Lithium;a plurality of solid Thorium fuel rods positioned within the tubular member and arranged such that at least a portion of each Thorium fuel rod is below the window opening in the lid, each solid Thorium fuel rod comprising:an inner member comprising Beryllium andan outer member formed from a solid that comprises at least some solid Thorium, wherein the outer member defines an opening passing through the solid Thorium fuel rod and the inner member is located within the opening;a plurality of immersion pumps, each immersion pump including:a shaft extending through one of the impeller openings, the shaft defining a first end coupled to a motor and a second end extending into the molten salt assembly, andan impeller coupled to the second end of the shaft,a primary heat exchange assembly comprising a first set of primary coils positioned within the main assembly body and a second set of primary coils positioned outside the main assembly body, the primary heat exchange assembly including an input pipe passing through the top lid and an output pipe passing through the top lid, the primary heat exchange assembly further including a non-Thorium containing molten salt within the primary coils and being configured such that the non-Thorium molten salt within the first set of primary coils is capable of absorbing heat generated in the main assembly body;a secondary heat exchange assembly comprising a set of secondary coils positioned outside the main assembly body, the set of secondary coils containing a fluid, the secondary coils being configured such that they are capable of absorbing heat from the second set of primary coils and generating vapor;a turbine-based electric generator assembly comprising a turbine and an electric generator driven by the turbine, the turbine being arranged such that it is capable of being driven by vapor generated by the secondary heat exchange assembly, the turbine-based electric generator further including a primary power output for the provision of useful electric power and a secondary power output, wherein the secondary power output of the turbine-based electric generator assembly is coupled to the power input of the proton beam source. 2. The Thorium molten salt energy system of claim 1 wherein the proton beam source includes a plurality of individual vacuum voltage chambers, wherein each individual vacuum chamber is coupled to a source of electrical power such that the interior of the vacuum chamber is maintained at a relatively uniform electric potential level, and wherein energization of a first group of the individual vacuum chambers provides protons at the first energy level and wherein energization of a subset of the vacuum chambers within the first group of the individual vacuum chambers provides protons at the second energy level. 3. The Thorium molten salt energy system of claim 2 wherein the first energy level is above 4.0 MeV. 4. The Thorium molten salt energy system of claim 3 wherein the second energy level is above 2.0 MeV. 5. The Thorium molten salt energy system of claim 1 wherein the proton beam source includes a quadrupole assembly capable of adjusting the shape of the proton beam. 6. The Thorium molten salt energy system of claim 5 wherein the quadrupole assembly is capable of adjusting the shape of the proton beam such that the beam can vary from a first ring shape of having a first inner diameter and a first outer diameter to a second ring shape having a second inner diameter and a second outer diameter, where the second inner diameter is greater than the first inner diameter. 7. The Thorium molten salt energy system of claim 6 wherein a first group of the solid Thorium fuel rods are positioned within the main assembly body such that they are within the first ring shape and a second group of solid Thorium fuel rods are positioned within the main body assembly such that they are within the second beam shape. 8. The Thorium molten salt energy system of claim 5 wherein the quadrupole assembly is capable of adjusting the shape of the proton beam from a first spot shape having a first diameter to a second spot shape having a second diameter, where the second diameter is greater than the first diameter. 9. The Thorium molten salt energy system of claim 1 wherein a single Thorium fuel rod is positioned within the main body assembly such that it is within the first spot shape and a plurality of Thorium fuel rods are positioned within the main body such that they are within the second spot shape. 10. The Thorium molten salt assembly of claim 1, further including high temperature seals positioned within each of the first group of at least two openings that are configured to prevent leakage of materials and gases from the interior of the main assembly body. 11. The Thorium molten salt assembly of claim 1 wherein each of the immersion pumps is an impeller pump driven by a variable speed motor. 12. The Thorium molten salt assembly of claim 1 wherein a space containing volume is defined by the upper surface of the molten salt containing Thorium and Lithium is such that an open region, not including molten salt but capable of containing gases, exists between the upper surface and the lower surface of the top lid. 13. The Thorium molten salt assembly of claim 1 wherein the open space includes a noble gas. 14. The Thorium molten salt assembly of claim 1 wherein at least one of the inner members comprising Beryllium defines an interior cavity that, for at least a plurality of points along the inner member, defines at least one Beryllium-containing projection upon which protons may impinge and at least one section through which protons may pass. 15. The Thorium molten salt assembly of claim 14 wherein the length of the at least one inner member comprising Beryllium is longer than the length of the outer member formed from a solid member that comprises at least some solid Thorium in which it is located. 16. The Thorium molten salt assembly of claim 9 wherein at least a portion of the inner member comprising Beryllium of the single Thorium fuel rod extends above the upper level of the molten salt within the main assembly body. 17. The Thorium molten salt energy system of claim 1 wherein the plurality of solid Thorium fuel rods are positioned between two support elements to form a modular Thorium fuel package. 18. The Thorium molten salt energy system of claim 17 wherein the modular Thorium fuel package comprises at least five solid Thorium fuel rods. 19. The Thorium molten salt energy system of claim 1 wherein the molten salt solution further comprises Beryllium. 20. The Thorium molten salt energy system of claim 1 wherein the non-molten salt fluid is water.
	description	The invention will now be illustrated by some examples, without however being limited to the compositions shown in the examples. A resin mixture based on a vinyl ester (Atlac XP 810, DSM Composite Resins, Zwolle, the Netherlands) with the following composition was prepared: The amount of resin mixture was split in 2 parts, and, using a doctor blade, spread to form a resin film on 2 polyethylenelpolyamide carrier films. The split, continuous carbon fibre bundles with a styrene-soluble binder (K value 80, split into 10 packages of 8K on average) from 14 bobbins were introduced into the chopper while spaced 4 cm apart above the bottom resin layer. The carbon fibre bundles were cut with rotating blades into packages of filaments approximately 2.5 cm long. On the bottom resin layer a 2D-randomly distributed and homogeneous fibre bed with a width of 50 cm was obtained. Next, the second resin layer was applied onto the fibre bed, after which the fibre material was impregnated with the resin mixture in a compacting unit. The sheet-shaped resin composition thus obtained was subsequently coiled. A carbon fibre loading of 52% by weight was achieved. Good wetting of the 8K packages with the resin mixture was distinctly visible up to the individual filaments. Thickening of the compound obtained took place in 10 days, after which the resin composition was moulded into a flat sheet using a mould coverage of 40%. A homogeneously black-coloured sheet was the result. Test specimens were sawn from the flat sheet obtained, and these were subsequently subjected to a bending test and an ILSS test. Results of bending test (ISO 178): Flexural modulus 33 GPa Flexural strength 500 MPa Results of ILSS test (ASTM 2344): 68 MPa The density of the material obtained was 1.39 g/cm3. Using the same resin mixture as in Example I and the same split, continuous carbon fibre bundle, a compound was made containing 25% by weight of carbon fibres with a length of approximately 2.5 cm. After thickening for 12 days a number of flat sheets were moulded with a mould coverage of 40% and with the following thicknesses 1.4 mm, 2.8 mm and 5.8 mm. The shielding efficiency was measured in accordance with ASTM 4935 in the frequency range between 30 MHz and 1 GHz. The shielding curves have a smooth profile across a wide frequency range. A number of characteristic values of these curves are shown in the table. Using the resin mixture of Example I, carbon fibre bundles from 14 bobbins with an unsplit 24K carbon fibre bundle were chopped to a length of approximately 2.5 cm above the bottom resin bed. The fibre bundles were spaced 4 cm apart as in Example I. Instead of a homogeneous fibre bed over a width of 50 cm, what was now obtained were rather 14 xe2x80x9cridgesxe2x80x9d of fibre material, in between which were areas where scarcely any fibre bundles were to be found. A carbon fibre loading of 48% by weight was achieved. To avoid such areas, the distance between the bundles in the chopper had to be reduced. Only when the bundles were spaced 1.5 cm apart could a homogeneous fibre distribution be achieved (the width of the SMC obtained then was approx. 20 cm). However, impregnation of this 20 cm wide fibre bed was found to be not well possible, as appeared from the high proportion of dry fibre bundles that were still present after the compacting unit. Using the resin mixture of Example I, split, continuous carbon fibre bundles with a non-styrene-soluble binder (48K, split into 7 packages of 7K on average) from 14 bobbins were cut to a length of approximately 2.5 cm. Under identical conditions to those in Example I, a 50 cm wide C-SMC was thus made with a carbon fibre loading of 52% by weight. After the compound had thickened for 10 days, a flat sheet was moulded using a mould coverage of 40%, from which a number of test bars were sawn for a bending test. On the surface of the flat sheet (and the test bars) areas were clearly perceptible where there were scarcely any carbon fibres (light-yellow in colour). As a result the strength of the material was relatively low. The results of the bending test according to ISO 178 were: Flexural strength: 200 MPaxc2x140 MPa Flexural modulus: 31 GPa. The ILSS (ASTM 2344) was 59 MPa. A resin mixture based on maleate resin (Palapreg 0423-N-2, DSM Composite Resins) was prepared, with the following composition: The amount of resin mixture was divided into 2 parts and, using a doctor knife, spread out to form a 30-cm wide resin film on 2 calender rolls. The split, continuous carbon fibre bundles with a styrene-soluble binder (K value 80, split into 10 packages of on average 8K) from 8 bobbins were introduced into the chopper above the calender gap while being spaced 4 cm apart. The carbon fibre bundles were cut with rotating knives into filament packages of approximately 2.5 cm length. The chopped fibres fell randomly distributed onto the resin layers on the rolls after which the fibre material together with the resin mixture passed through the gap between the 2 rolls and was impregnated with the resin mixture. The bulk resin composition thus obtained was removed from the rolls by means of scrapers and packed. A carbon fibre loading of 17% by weight was achieved. Good wetting of the 8K packages with the resin mixture was distinctly visible up to the individual filaments. Thickening of the obtained compound took place in 7 days, after which the resin composition was moulded into a flat sheet. A homogeneously dark-grey coloured sheet was the result. From the flat sheet obtained test specimens were sawn, which were subsequently subjected to a bending test. Results of bending test (ISO 178): Flexural modulus 20 GPa Flexural strength 200 MPa The density of the material obtained was 1.79 g/cm3. Using the resin mixture of Example III, carbon fibre bundles from 8 bobbins with a non-split 24K carbon fibre bundle were now cut above the calender to a length of approximately 2.5 cm. The distance between the fibre bundles was 4 cm, as in Example III. Instead of a homogeneous fibre distribution over a width of 30 cm, now rather 8 xe2x80x9cstripesxe2x80x9d of fibre material were obtained, with in between them areas where scarcely any fibre bundles were to be found. A carbon fibre loading of 16% by weight was achieved. To avoid the formation of such areas with and without fibre bundles, the distance between the bundles in the chopper had to be reduced. When the bundle spacing was smaller than 1.5 cm a more homogeneous fibre distribution could be achieved, but impregnation of the fibres was found to be not well possible, which appeared from the large proportion of dry fibre bundles that were still present in the bulk compound obtained. Example I was repeated without any LPA being present. The resin mixture used had the following composition: The same compounding procedure and split, continuous carbon fibre bundles as in Example I were applied. A carbon fibre loading of 52% by weight was achieved. Good wetting of the 8K packages with the resin mixture was distinctly visible up to the individual filaments. Thickening of the compound obtained took place in 10 days, after which the resin composition was moulded into a flat sheet using a mould coverage of 40%. A homogeneously black-coloured sheet was the result. Test specimens were sawn from the flat sheet obtained, and these were subsequently subjected to a bending test and an ILSS test. Results of bending test (ISO 178): Flexural modulus 33 gpa flexural strength 530 mpa Results of ILSS test (ASTM 2344): 68 mpa
	abstract	The present invention includes an apparatus for watertight sealing of a steam generator nozzle and methods for installing the apparatus. The apparatus comprises a nozzle dam, a nozzle dam attachment ring, and a seal. The attachment ring is provided in an interior of the nozzle and has a plurality of retaining tabs and a nozzle dam landing. The nozzle dam is adapted for insertion into the attachment ring and abutment against the nozzle dam landing. The nozzle dam has a plurality of radial protrusions adapted to interlock with the retaining tabs for fixing the nozzle dam in the attachment ring upon rotation of the nozzle dam in the attachment ring. The seal covers at least one side of the nozzle dam for effecting a seal between the nozzle dam and the attachment ring. The present invention also provides methods and apparatus for the pressurization and control of nozzle dam seals.
	claims	1. An Anti-Scatter-Grid for a radiation detector, comprising radiation absorbing lamellae which are adapted to produce a signal that indicates an amount of absorbed radiation. 2. The Anti-Scatter-Grid according to claim 1, wherein the lamellae comprise a semiconductor material which converts absorbed radiation into electrical signals, particularly a material with a low intrinsic energy conversion coefficient for conversion of photons into electron-hole pairs. 3. The Anti-Scatter-Grid according to claim 1, wherein the lamellae comprise a scintillator material for conversion of incident radiation of a first energy level into radiation of a second energy level. 4. The Anti-Scatter-Grid according to claim 1, wherein the lamellae comprise a material with a high absorption coefficient, particularly higher than 1 cm−1, for photons with energies below 150 keV. 5. The Anti-Scatter-Grid according to claim 1, wherein the lamellae are at least partially covered by electrodes. 6. The Anti-Scatter-Grid according to claim 5, wherein at least one of the electrodes ends at a distance away from an edge of the corresponding lamella, wherein said distance is preferably such that radiation traveling through the material of the lamella is substantially absorbed after said distance. 7. The Anti-Scatter-Grid according to claim 6, wherein it comprises both electrodes ending at a distance away from an edge and electrodes ending at the edge of the corresponding lamella. 8. A radiation detector, comprisingan Anti-Scatter-Grid according to claim 1;optionally a converter for conversion of incident radiation of a first energy level into radiation of a different energy level;an array of radiation sensitive sensor units;a signal processing unit for evaluation of signals generated by the Anti-Scatter-Grid. 9. The radiation detector according to claim 8, wherein the signal processing unit is adapted to discriminate fractions of incident radiation with respect to their parallelism to channels and/or lamellae of the Anti-Scatter-Grid. 10. An examination apparatus with an imaging system, particularly an X-ray, CT, PET, SPECT or nuclear imaging device, the imaging system comprising an X-ray sensitive radiation detector according to claim 8.
	abstract	A fabrication method of burnable absorber nuclear fuel pellets and burnable absorber nuclear fuel pellets fabricated by the same are provided, in which the fabrication method includes adding boron compound and manganese compound to one or more type of nuclear fuel powders selected from the group consisting of uranium dioxide (UO2), plutonium dioxide (PuO2) and thorium dioxide (ThO2) and mixing the same (step 1), compacting the mixed powder of step 1 into compacts (step 2), and sintering the compacts of step 2 under hydrogen atmosphere (step 3). According to the fabrication method, sintering can be performed under hydrogen atmosphere at a temperature lower than the hydrogen atmosphere sintering that is conventionally applied in the nuclear fuel sintered pellet mass production, by adding sintering additives such as manganese oxide or the like.
052276830	claims	1. A permanent magnet assembly for treating a fluid flowing inside a pipe having a pipe wall of predetermined dimensions, comprising: a pair of magnetic devices, each device comprising: said pair of magnetic devices each being shaped and sized so that when each device is assembled with said magnet sandwiched between said pole pieces with said proximal ends of said pole pieces abutting said respective oppositely facing surfaces of said magnet, said devices can be placed on opposite sides of said pipe at corresponding positions, with said distal ends of said pole pieces of each device longitudinally spaced along said pipe, with opposite poles of said respective devices facing each other, a magnetic field will extend between each pair of said facing opposite poles, across said pipe and transversely to the direction of fluid flow therein so that said fluid, flowing inside said pipe will cut said flux lines to generate an electrical current in said fluid so as to reduce scaling, corrosion, and algae buildup in said pipe. (a) said permanent magnet of each of said magnetic devices produces a flux density of about 12,000 gauss, and wherein (b) the cross-sectional area of each of said distal ends of each of said pole pieces is smaller than said oppositely facing surfaces of said permanent magnet, whereby said pole pieces will condense the magnetic fields of said magnets, thereby magnetically saturating said pole pieces and subsequently oversaturating said pipe wall, driving magnetic flux through said pipe wall. (a) the thickness of said permanent magnets and the lengths of said pole pieces is such that the spacing between said distal ends of said pole pieces is larger than the inside diameter of said pipe and wherein (b) said distal end of each of said pole pieces has a thickness, measured in the longitudinal direction of said pipe, greater than the thickness of the wall of said pipe. (a) the cross-sectional area of each of said distal ends of each of said pole pieces is smaller than said oppositely facing surfaces of said permanent magnet, whereby said pole pieces will condense the magnetic fields of said magnets, thereby magnetically saturating said pole pieces and subsequently oversaturating said pipe wall, driving magnetic flux through said pipe wall, and wherein (b) the thickness of said permanent magnets and the lengths of said pole pieces is such that the spacing between said distal ends of said pole pieces is larger than the inside diameter of said pipe. 2. The assembly of claim 1 wherein said permanent magnet of each of said magnetic devices produces a flux density of about 12,000 gauss. 3. The assembly of claim 1 wherein the cross-sectional area of each of said distal ends of each of said pole pieces is smaller than said oppositely facing surfaces of said permanent magnet, whereby said pole pieces will condense the magnetic fields of said magnets, thereby magnetically saturating said pole pieces and subsequently oversaturating said pipe wall, driving magnetic flux through said pipe wall. 4. The assembly of claim 3 wherein the cross-sectional area of each of said distal ends of each of said pole pieces is approximately 1.5 times smaller than said oppositely facing surfaces of said permanent magnet, whereby said pole pieces will condense the magnetic fields of said magnets. 5. The assembly of claim 1 wherein the thickness of said permanent magnets and the lengths of said pole pieces is such that the spacing between said distal ends of said pole pieces is larger than the inside diameter of said pipe. 6. The assembly of claim 5 wherein the thickness of said permanent magnets and the lengths of said pole pieces is such that the spacing between said distal ends of said pole pieces is approximately 1.25 times larger than the inside diameter of said pipe. 7. The assembly of claim 1 wherein said distal end of each of said pole pieces has a thickness, measured in the longitudinal direction of said pipe, greater than the thickness of the wall of said pipe. 8. The assembly of claim 7 wherein said distal end of each of said pole pieces has a thickness, measured in the longitudinal direction of said pipe, approximately three times greater than the thickness of the wall of said pipe. 9. The assembly of claim 1, further including said pipe, said pipe having said predetermined dimensions. 10. The assembly of claim 9 wherein said pipe is made of steel. 11. The assembly of claim 1 wherein 12. The assembly of claim 11 wherein the cross-sectional area of each of said distal ends of each of said pole pieces is approximately 1.5 times smaller than said oppositely facing surfaces of said permanent magnet, whereby said pole pieces will condense the magnetic fields of said magnets. 13. The assembly of claim 1 wherein 14. The assembly of claim 13 wherein said distal end of each of said pole pieces has a thickness, measured in the longitudinal direction of said pipe, approximately three times greater than the thickness of the wall of said pipe. 15. The assembly of claim 14, further including said pipe, said pipe having said predetermined dimensions. 16. The assembly of claim 15 wherein said pipe is made of steel. 17. The assembly of claim 16 wherein said permanent magnet of each of said magnetic devices produces a flux density of about 12,000 gauss. 18. The assembly of claim 1 wherein 19. The assembly of claim 18 wherein said distal end of each of said pole pieces has a thickness, measured in the longitudinal direction of said pipe, greater than the thickness of the wall of said pipe. 20. The assembly of claim 19, further including said pipe, said pipe having said predetermined dimensions.
	claims	1. A method for operating a nuclear reactor in order to produce electricity, the reactor comprising a core that is loaded with assemblies that comprise nuclear fuel rods, at least one nuclear fuel rod being of a type comprising:a cladding of recrystallized zirconium-based alloy, comprising, by mass,from 0.8 to 1.3% of niobium,between 1000 ppm and 1700 ppm of oxygen,between 0 and 35 ppm of sulphur,between 0 and 7000 ppm in total of at least one of iron, chromium and vanadium,between 0 and 2% of tin,between 0 and 70 ppm of nickel,between 0 and 100 ppm of carbon, andbetween 0 and 50 ppm of silicon,a balance being constituted by zirconium, with an exception of inevitable impurities, andpellets of nuclear fuel based on uranium oxide, the pellets being stacked inside the cladding, the method comprising:controlling operation of the reactor, during a class 2 transient power occurrence, such that at least one of:(a) a linear power density of the nuclear fuel rod remains lower than a limit linear power density, the limit linear power density being greater than 430 W/cm and less than or equal to 444 W/cm, and(b) a variation of the linear power density of the nuclear fuel rod remains lower than a limit variation, the limit variation being greater than 180 W/cm and less than or equal to 253 W/cm. 2. The method as recited in claim 1 further comprising initiating a corrective action or activating an alarm when the linear power density exceeds the limit linear power density or the variation of the linear power density exceeds the limit variation. 3. The method according to claim 1, wherein the limit linear power density is greater than 440 W/cm. 4. The method according to claim 1, wherein the limit variation is greater than 200 W/cm. 5. The method according to claim 4, wherein the limit variation is greater than 220 W/cm. 6. The method according to claim 1, wherein the alloy further comprises:between 5 and 35 ppm of sulphur by mass. 7. The method according to claim 1, wherein the alloy comprises between 0.03 and 0.25% in total of at least one of iron, chromium and vanadium. 8. The method according to claim 1, wherein the recrystallized zirconium alloy has been subjected to at least one annealing operations at temperatures of less than 600° C. 9. The method according to claim 1, wherein an inner side of the cladding is pressurized, before use, to a pressure of less than 20 bar. 10. The method as recited in claim 1 wherein the controlling includes using a deformation energy density for ensuring that the linear power density of the nuclear fuel rod remains lower than the limit linear power density, or the variation of the linear power density of the nuclear fuel rod remains lower than the limit variation. 11. The method according to claim 1, wherein the pellets include the metal oxide Cr2O3. 12. The method according to claim 11, wherein the pellets comprise from 1200 to 2000 ppm by mass of Cr2O3. 13. The method according to claim 12, wherein the pellets comprise from 1450 to 1750 ppm by mass Cr2O3. 14. A method for operating a nuclear reactor in order to produce electricity, the reactor comprising a core that is loaded with assemblies that comprise nuclear fuel rods, at least one nuclear fuel rod being of a type comprising:a cladding of recrystallized zirconium-based alloy, comprising, by mass,from 0.8 to 1.3% of niobium,between 1000 ppm and 1700 ppm of oxygen,between 0 and 35 ppm of sulphur,between 0 and 7000 ppm in total of at least one of iron, chromium and vanadium,between 0 and 2% of tin,between 0 and 70 ppm of nickel,between 0 and 100 ppm of carbon, andbetween 0 and 50 ppm of silicon,a balance being constituted by zirconium, with an exception of inevitable impurities, andpellets of nuclear fuel based on uranium oxide, the pellets being stacked inside the cladding, the pellets including at least one metal oxide for increasing a thermal creep of the pellets, the method comprising:controlling operation of the reactor, during a class 2 transient power occurrence, such that at least one of:(a) a linear power density of the nuclear fuel rod remains lower than a limit linear power density, the limit linear power density being greater than 430W/cm and less than or equal to 620 W/cm, and(b) a variation of the linear power density of the nuclear fuel rod remains lower than a limit variation, the limit variation being greater than 180 W/cm and less than or equal to 450 W/cm. 15. The method as recited in claim 14 further comprising initiating a corrective action or activating an alarm when the linear power density exceeds the limit linear power density or the variation of the linear power density exceeds the limit variation. 16. The method as recited in claim 14, wherein the limit linear power density is greater than 440 W/cm. 17. The method as recited in claim 14, wherein the limit variation is greater than 200 W/cm. 18. The method as recited in claim 14, wherein the limit variation is greater than 220 W/cm. 19. The method as recited in claim 14, wherein the alloy further comprises:between 5 and 35 ppm of sulphur by mass. 20. The method as recited in claim 14, wherein the alloy comprises between 0.03 and 0.25% in total of at least one of iron, chromium and vanadium. 21. The method as recited in claim 14, wherein the recrystallized zirconium alloy has been subjected to at least one annealing operations at temperatures of less than 600° C. 22. The method as recited in claim 14, wherein the limit linear power density is greater than 590 W/cm. 23. The method as recited in claim 14, wherein the limit linear power density is greater than 600 W/cm. 24. The method as recited in claim 14, wherein the limit linear power density is greater than 610 W/cm. 25. The method as recited in claim 14, wherein the limit variation is greater than 430 W/cm. 26. The method as recited in claim 14, wherein the limit variation is greater than 440 W/cm. 27. The method as recited in claim 14, wherein the metal oxide is Cr2O3. 28. The method as recited in claim 27, wherein the pellets comprise from 1200 to 2000 ppm by mass of Cr2O3. 29. The method as recited in claim 28, wherein the pellets comprise from 1450 to 1750 ppm by mass Cr2O3. 30. The method as recited in claim 14, wherein an inner side of the cladding is pressurized, before use, to a pressure of less than 20 bar. 31. The method as recited in claim 17 wherein the controlling includes using a deformation energy density for ensuring that the linear power density of the nuclear fuel rod remains lower than the limit linear power density, or the variation of the linear power density of the nuclear fuel rod remains lower than the limit variation.
	abstract	Disclosed herein is an apparatus comprising: a radiation absorption layer comprising an electrode; a counter configured to register a number of radiation particles absorbed by the radiation absorption layer; a controller configured to start a time delay from a time at which an absolute value of an electrical signal on the electrode equals or exceeds an absolute value of a first threshold; a comparator configured to compare the electrical signal to a second threshold; wherein the controller is configured to activate the comparator during the time delay; wherein the controller is configured to cause the number registered by the counter to change, if the comparator determines that an absolute value of the electrical signal equals or exceeds an absolute value of the second threshold.
049884751	abstract	The device comprises a rod (28) on which is mounted a means (30, 32, 33) for support and displacement which is movable in an axial direction of the rod (28) and in two directions perpendicular to this axial direction. The support (33) of the means for displacement carries a device for pushing axially on a longitudinal end of the pencil, which may be activated by a control means (35), one end of which is located in a control station (25) disposed above a fuel assembly storage pool. Means for measuring the axial pushing force on the pencil and the length of axial displacement of the pencil under the effect of the pushing are disposed in the control station (25). A video camera (34), carried by the support (33), makes it possible to provide an image of a zone in the vicinity of the end of the pencil.
062696297	summary	BACKGROUND OF THE INVENTION There is a need for improved plasma thrusters for attitude control, station-keeping, and primary propulsion of very small satellites in space such as "Microsats" weighing less than 20 kg and "Nanosats" weighing less than one kg. These small satellites are expected to be widely used for Air Force and commercial applications. Such attitude control thrusters should be packaged in small lightweight containers and be highly efficient so as to employ small amounts of energy, typically about one watt. Relative to standard sized satellite thrusters, such small pulsed plasma thrusters (PPT), having 10-100 size reductions, should eliminate the prior art PPT which require separate spark plug igniter devices, in favor of lightweight components permitting precise control of a number of small impulse propellant modules, by means of a single electrical pulser for selectively actuating desired propellant impulse producing modules via a plurality of transmission lines. It is desired to provide a reduction of attitude control system (ACS) thruster mass by ninety percent or more, obtained by eliminating conventional torque rods or reaction wheels, in accordance with the placement of lightweight low cost propellant modules at the corners of the spacecraft, each module independently and selectively controlled by a single low mass centrally located pulse generating thruster control device which can employ an energy storage capacitor. Also, the arrangement has none of the aforesaid prior art spark plug igniters. The use of standard, readily available components, such as those previously flight qualified, is also desired. SUMMARY A PREFERRED EMBODIMENT OF THE INVENTION The aforesaid needs are met by providing a precisely controlled microPPT, e.g. one watt, producing small reproducible impulse bits of thrust by means of a plurality of coaxial cable propellant modules, providing increased thrust for a small fixed discharge energy, which reduces the volume required to store the propellant mass and which decreases radiated electromagnetic interference (EMI) relative to prior thrusters. More specifically, each propellant module, which can be positioned at a selected outer portion of the satellite, comprises a semi-rigid section of an ordinary low cost coaxial cable. High voltage pulses, initiated by a charged energy storage capacitor of a thruster control pulse generator, are selectively impressed or discharged across the outer conductive cylindrical wall of an ordinary coaxial cable segment of each propellant module and its inner central conductor, to produce vaporization of the solid "TEFLON" polymer spacer or filler, separating the inner and outer conductors, and in turn generating small precise impulse forces oriented to permit easy attitude satellite control. The result is a lightweight PPT having no sparkplug igniter, and using significantly less propellant to perform a given small satellite maneuver than alternative designs such as cold gas thrusters and those using thermal acceleration. Changing the capacitor discharge frequency and thus the frequency of the resulting pulse train applied across the coax components can also be used to change the degree of thrust.
	description	The present application is the national phase of International Application No. PCT/CN2015/090250, entitled “MULTI-LAYER STAGGERED COUPLING COLLIMATOR, RADIATOR, DETECTOR AND SCANNER” filed on Sep. 22, 2015, which claims priority to Chinese Patent Application No. 201410512964.4, titled “MULTI-LAYER STAGGERED COUPLING COLLIMATOR, RADIATOR, DETECTION DEVICE AND SCANNING APPARATUS”, filed on Sep. 29, 2014 with the State Intellectual Property Office of People's Republic of China, both of which are incorporated herein by reference in their entireties. 1. Field of the Disclosure The present disclosure relates to the technical field of nuclear medicine imaging, and in particular to a multi-layer staggered coupling collimator, a radiator, a detection device and a scanning apparatus. 2. Discussion of the Background Art A Single Photon Emission Computed Tomograph (abbreviated as SPECT hereinafter) is an advanced nuclear medicine molecular imaging tool, which can obtain metabolism information on an organism in a noninvasive manner, and play a major role in mechanism study, diagnosis and treatment of critical diseases such as a cardiovascular system disease, a nervous system disease and tumor. The SPECT includes a probe, a rotatable rack, a scanning table, and an image acquisition and processing workstation, where the probe generally includes a scintillator detector and a collimator. In imaging, drugs with radioactive nuclides such as Tc-99m are injected into a human body, and the probe is arranged around the human body to acquire gamma rays from different angles so as to obtain two dimensional radioactive intensity distribution maps from different angles. Then, a three dimensional image that reflects radioactive drug distribution of the human body is obtained by image reconstruction. A spatial resolution, sensitivity and an imaging field of view are three most important performance indexes of the SPECT. The spatial resolution reflects a capability of the SPECT to distinguish details of an object, and improvement of the spatial resolution can increase richness and definition of details of the acquired image. The sensitivity reflects a capability of the SPECT to detect an object with a low activity, and improvement of the sensitivity can reduce an injection amount of the radioactive drug or imaging time. The imaging field of view reflects a range of sizes of objects which can be scanned by the SPECT, and improvement of the SPECT can accelerate or optimize scanning on a big object. Improvement of one or more of the indexes is a foundational direction of development of the SPECT over years. The performance of the collimator is one of the main factors influencing the performance of the SPECT system. Optimizing the design of the collimator is an important way to improve the performance of the SPECT system. The collimator is generally a square plate in which through-holes are arranged densely. The plate is generally made of a heavy metal such as lead and tungsten or an alloy thereof, which can block gamma photons that do not fly in holes and allow gamma photons flying through the holes. A linear track of flight of gamma photons can be determined based on the direction of the holes on the collimator and positions on the detector radiated by the gamma photons obtained by the scintillator detector. Performances of the collimator are generally also indicated by the spatial resolution, the sensitivity and the field of view and the like, and the performance indexes are determined by geometric parameters (such as a size of the plate, a shape, a size and a deepness of a hole), a material and a machining precision of the collimator. The indexes are in a mutual constraint relation, that is, increase of one index normally results in decrease of another index. Generally, a suitable combination of performances is selected by performing optimization according to an application demand. Presently, for imaging of a big animal or a person, the resolution and the sensitivity of the SPECT system needs to be improved greatly. Compared with a positron emission computed tomography which is also a nuclear medicine molecular imaging tool, the SPECT system has a lower resolution and an even lower sensitivity by two orders of magnitude. The low sensitivity and resolution limit the capability of the SPECT system (particularly a clinic SPECT system based on a parallel hole collimator) to diagnose diseases early, find and analyze a lesion quantitatively, and reflect details of an object accurately. Therefore, designing a high performance collimator is still effective means to improve the performances of the SPECT system. In actual imaging with the clinic SPECT, collimators with different performances may be selected according to application conditions. For this reason, it needs to dismount an old collimator and install a new collimator. Since the collimator is big and heavy, it is not convenient to replace the collimator, and damages may occur during the replacing process. In addition, an SPECT system is normally configured with a very small number of collimators with different performances, and therefore a range of selection is small, which means there may not be sufficient options of collimators to be selected for different applications. An ideal solution to this problem is to design a multiple-performance collimator having adjustable performances. Processing difficulty is also a main factor to be considered in designing the collimator, and a collimator with an extreme performance or a collimator with complex functions may be impossible due to processing difficulty. Therefore, it needs to consider processing difficulty of the collimator when designing the collimator, in addition to functions of the collimator. Therefore, it is desired to provide a multi-layer staggered coupling collimator with an improved structure, a radiator, a detection device and a scanning apparatus, so as to solve the problems in the conventional technology. In view of above, an object of the present disclosure is to provide a multi-layer staggered coupling collimator, a radiator, a detecting device and a scanning apparatus, to improve one or more performance indexes of the collimator, or support multiple performance combinations to be adjusted or selected. In order to achieve the above object, the following technical solutions are provided according to the present disclosure. A multi-layer staggered coupling collimator is provided, which includes multiple collimation layers, where multiple collimation holes are provided on each of the multiple collimation layers and at least two of the multiple collimation layers are coupled to each other in a staggered manner. A radiator is provided, which includes a radiation source and the multi-layer staggered coupling collimator described above, and the multi-layer staggered coupling collimator is configured to collimate rays emitted by the radiation source. A detection device is provided, which includes a detector and the multi-layer staggered coupling collimator described above, and the multi-layer staggered coupling collimator is configured to collimate rays, where the collimated rays are applied to the detector. A scanning apparatus is provided, which includes a detection device and a rack, where the detection device is installed on the rack, the detection device includes the multi-layer staggered coupling collimator described above, and the multi-layer staggered coupling collimator is configured to collimate rays. It can be seen from the above technical solutions that, compared with a single layer collimator, the multi-layer staggered coupling collimator according to the present disclosure not only has improved performances and multiple selectable performances, but also has good machinability. Since the collimator includes multiple collimation layers, each of the multiple collimation layers is thin, which ensures machining precision. Before the present disclosure is introduced in detail, concepts of “a collimation layer” and “staggered coupling” are defined first. The collimation layer may be understood from the following two aspects. In a first aspect, the collimation layer may be regarded as a sub-collimator obtained by dividing a collimator (shortening a hole). In a second aspect, the collimation layer may be regarded as one of multiple collimators forming a collimation system by superimposing in series. When a gamma photon passes through a collimator including multiple collimation layers, the gamma photon needs to pass through all collimation layers in sequence. The “staggered coupling” means that center lines of corresponding collimation holes of two collimation layers are not aligned when the two collimation layers are coupled in series. In contrast, alignment coupling means that center lines of corresponding collimation holes of two collimation layers are aligned. According to the present disclosure, a multi-layer staggered coupling collimator, a radiator, a detection device and a scanning apparatus are provided, which can improve one or more performance indexes of the collimator or can support multiple performance combinations to be adjusted or selected. According to the present disclosure, a radiator is further provided, which includes a radiation source and the multi-layer staggered coupling collimator according to the present disclosure. The multi-layer staggered coupling collimator is configured to collimate rays generated by the radiation source. According to the present disclosure, a detection device is further provided, which includes a detector and the multi-layer staggered coupling collimator according to the present disclosure. The multi-layer staggered coupling collimator is configured to collimate rays, where the collimated rays are applied to the detector. According to the present disclosure, a scanning apparatus is further provided, which includes a detection device and a rack, where the detection device is installed on the rack, and the detection device includes the multi-layer staggered coupling collimator according to the present disclosure. The multi-layer staggered coupling collimator is configured to collimate rays. The multi-layer staggered coupling collimator disclosed in multiple embodiments in the present disclosure is provided in each of the radiator, the detection device and the scanning apparatus disclosed by the present disclosure. Hereinafter a structure of the multi-layer staggered coupling collimator included in the radiator, the detection device and the scanning apparatus is described in detail. As shown in FIG. 1 and FIG. 2, the multi-layer staggered coupling collimator includes multiple collimation layers. Multiple collimation holes are provided on each of the multiple collimation layers. At least two of the multiple collimation layers are coupled to each other in a staggered manner. Collimation holes of the collimation layers may have the same type or different types. Generally the collimation holes have the same type, while the type of the collimation hole is selected according to actual application demands. The staggered coupling collimator includes three types in the following. In a first type, positions of all the collimation layers of the staggered coupling collimator are unchangeable. That is, after the collimation layers are staggered, all the collimation layers can not move any more relative to each other. The collimation layer may be fixed by means of screw in a hole punched in the collimation layer, or by means of a housing with a slot. The multi-player staggered coupling collimator of the first type has only one manner of staggering, but performance indexes of the collimator can still be improved. In a second type, positions of a part of the collimation layers of the staggered coupling collimator are adjustable. That is, after the collimation layers are staggered, the collimation layers may be staggered again according to a new demand. Not all the collimation layers are changeable. Only positions of a part of the collimation layers are changeable to be staggered. The collimator is provided with a guide rail, the part of the changeable collimation layers may be adjusted with the guide rail. The guide rail may have a function of lock catch. Performance indexes of the collimator can be improved after staggering. In a third type, positions of all the collimation layers of the staggered coupling collimator are adjustable. That is, after the collimation layers are staggered, the collimation layers may be staggered again according to a new demand. Any of the collimation layers may be selected to be staggered as needed. The adjustable collimation layers may be adjusted with a guide rail as in the second type. The guide rail may have a function of lock latch. Performance indexes of the collimator can be improved after staggering. The multi-layer staggered coupling collimator provided according to the present disclosure has two features: firstly, gamma beams are collimated by at least two collimation layers in the collimator, and secondly, at least two collimation layers are coupled to each other in a staggered manner. The second feature includes two configurations: firstly, all adjacent collimation layers are coupled to each other in a staggered manner; and secondly, not all adjacent collimation layers are coupled to each other in a staggered manner. The second configuration is mainly applied to a case of a large number of collimation layers. For example, a first layer and a second layer are aligned and not staggered, while the first layer and a third layer are staggered. Apparently, since the first layer and the second layer are aligned, the second layer and third layer are staggered. In a case that the collimator includes multiple collimation layers, thicknesses of the multiple collimation layers may be different, or may be the same. A thickness of each layer may be determined according to an actual demand. In a case that the thicknesses of the collimation layers are different, a ratio of a minimum thickness to a maximum thickness ranges from 1:1 to 1:11. An effective aperture size and even a shape of the collimation holes in a case of staggered coupling are different from that in a case of alignment coupling, resulting in different performances between the staggered coupling collimator and the alignment coupling collimator (equivalent to a single layer collimator). Further, an internal structure and a performance of the multi-layer staggered coupling collimator are changed as a staggering direction, a staggering amount, the number of layers of the collimator or a layer thickness is changed. Based on this principle, with the multi-layer staggered coupling collimator according to the present disclosure, a high performance collimator can be implemented in which at least one of a spatial resolution and sensitivity is improved (compared with a single layer collimator equivalent to a collimator for which all layers are coupled to each other in an alignment manner), and a performance adjustable collimator can be implemented in which one or more of parameters such as a staggering direction for the layers, a staggering amount for the layers, the number of layers, a thickness of the layers are adjustable. The multi-layer staggered coupling collimator according to the present disclosure may include N collimation layers coupled in a staggered manner, where a value of N may range from 2 to 30, or may be greater than 30. The value of N is selected according to an actual application demand, in combination with the number and thickness of layers, assembling ability of the collimator and the like. At least two collimation layers among the N collimation layers are coupled in a staggered manner. Practically, it may be the case that all adjacent collimation layers are coupled in a staggered manner. In this case, a position of the first layer may be the same as a position of the third layer. Optional collimation layer types include a type of parallel hole collimator, a type of divergent collimator and a type of convergent collimator. Different collimation layer types have different structure features, different performances, and different applications. The parallel hole collimator has the features that all through-holes are parallel, such that an object's image has the same size as the object. This collimator is most commonly used in a clinical SPECT requiring an equal sensibility and resolution in a large field of view. The divergent collimator has the features that holes arranged densely on the collimator are not parallel, such that a pitch of holes becomes larger from a gamma camera end to an object end, so as to generate a shrunken image of the object. This collimator may be applied to a system for imaging a large object with a small detector. The convergent collimator has the reverse feature to the divergent collimator, that a pitch of holes becomes smaller from a gamma camera end to an object end, so as to generate an amplified image of the object. This collimator is generally applied to a local imaging application requiring a high resolution and high sensitivity in a small field of view. A pinhole collimator has the feature that only one small hole is used to generate an inverted image of an object based on a pinhole imaging principle. This collimator is generally applied to a SPECT system requiring a high resolution in a small field of view, and is commonly used in animal imaging. Reference is made to FIG. 8, which is a schematic diagram of a convergent dual layer staggered coupling collimator. A detector is placed on a right side of the collimator, i.e., a right side of FIG. 8. The collimator is convergent in relative to the detector. In a case that rays are emitted from a left side to a right side, the collimator is convergent; and in a case that rays are emitted from the right side to the left side, the collimator is divergent. Convergence or divergence is defined in relative to the detector. Types and staggering design of collimation holes of the multi-layer staggered coupling collimator according to the present disclosure include but not limited to the following five designs. In a first design, as shown in FIG. 2, the collimator has the following features. Collimation holes on the collimation layers have a shape of square (for a clear description, it is assumed that a direction of one group of opposite sides of the square is y direction and a direction of the other group of opposite sides is z direction). The collimation holes are arranged in a square grid (arranged in the y direction and the z direction); all adjacent collimation layers among the N collimation layers are staggered in only the y direction, in only the z direction, or in both the y direction and the z direction. The collimation layers are staggered such that a pitch of holes of a squared pattern obtained by projecting the collimator in a direction parallel to the hole direction is ½ to 1/M of a pitch of holes on the collimation layers, where a value of M ranges from 2 to N. In a second design, as shown in FIG. 3 and FIG. 4, the collimator has the following features. Collimation holes on a collimation layer have a shape of regular hexagon, and the regular hexagon holes are arranged in a regular triangular grid (each grid unit has a shape of regular triangular, and each grid point corresponds to a center of a collimation hole). All adjacent collimation layers among the N collimation layers are staggered with each other. In the staggered collimation layers, a center of a collimation hole on a collimation layer is aligned with a common vertex of adjacent hexagon holes of an adjacent collimation layer (distances between the common vertex and centers of three adjacent holes are the same). In a third design, as shown in FIG. 6, the collimator has the following features. Collimation holes on a collimation layer have a shape of round or any polygon approximate to round. The holes with the shape of round or any polygon approximate to round are arranged in a regular triangular grid. All adjacent collimation layers among the N collimation layers are staggered with each other. In the staggered collimation layers, a center of a hole on a collimation layer is aligned with a point on an adjacent collimation layer (the point is located in a region surrounded by three holes which are adjacent to each other, and distances between the point and centers of the three holes are the same). In a fourth design, as shown in FIG. 6, the collimator has the following features. Collimation holes on a collimation layer have a shape of round or any polygon approximate to round, and the collimation holes are arranged in a square gird (each grid unit is square, and each grid point in the grid corresponds to a center of a collimation hole); and all adjacent collimation layers among the N collimation layers are staggered. Assuming that a direction of one group of opposite sides of the square is y direction and a direction of the other group of opposite sides is z direction, the collimation layers are staggered in the y direction and/or z direction, i.e., in a direction of only one group of parallel sides of the square grid or in both directions of the two group of parallel sides of the square grid. The staggering amount is ½ of a pitch of holes. In a fifth design, as shown in FIG. 5, the collimator has the following features. Collimation holes on a collimation layer have a shape of regular triangle, and the collimation holes are arranged in a regular hexagon grid (each grid unit is a regular hexagon, and each grid point in the grid corresponds to a center of a collimation hole). All adjacent collimation layers among the N collimation layers are staggered. The collimation layers are staggered in a direction of one side of a triangular hole by a staggering amount of sqrt(3)/2 times of a side length of a regular hexagon grid unit (sqrt indicates an extraction operation). An approximate value may be assigned to the staggering amount in a case that the staggering amount is an infinite non-circulating decimal or an infinite circulating decimal. As shown in FIG. 7, the collimators in FIG. 7(a), FIG. 7(b) and FIG. 7(c) have the same total thicknesses and the same sizes and pitches of holes on collimation layers, while staggering design is adopted in FIG. 7(b) and FIG. 7(c), such that thicker beams (indicated by dotted lines and arcs in the figures) can pass, leading to higher sensitivity than the single layer design in FIG. 7(a). In FIG. 7(b), a ratio of thicknesses of two collimation layers is 1:1, and in FIG. 7(c), a ratio of thicknesses of two collimation layers is 3:1. Although the design of two layers is adopted in both FIG. 7(b) and FIG. 7(c), they have different effect because of the different ratios of thicknesses. In FIG. 7(b), a thicker beam from a point source is allowed to pass, leading to higher sensitivity. As shown in FIG. 9, collimators shown in FIG. 9(a) and FIG. 9(b) have a same total thickness, in which adjacent layers are coupled in a staggered manner. In FIG. 9(a), three collimation layers are provided, and a ratio of thicknesses of the three collimation layers is 1:1:2 from left to right. In FIG. 9(b), five collimation layers are provided, and a ratio of thicknesses of the five collimation layers is 1:1:1:1:2 from left to right. Both of the two designs have a higher resolution than the single layer design. Because of the different numbers of layers, a higher resolution can be obtained in FIG. 9(b) than in FIG. 9(a). As shown in FIG. 10, FIG. 10(a), FIG. 10(c), FIG. 10(e) and FIG. 10(g) are schematic diagrams of a first design. FIG. 10(a) shows three beams (each beam is indicated by dotted-line boundaries and an arc) emitted from one point which can pass through a collimator. FIG. 10(c) shows first parallel beams which can pass through the collimator (emitting towards lower right). FIG. 10(e) shows second parallel beams which can pass through the collimator (emitting in a direction parallel to the holes). FIG. 10(g) shows third parallel beams which can pass through the collimator (emitting towards upper right). FIG. 10(b), FIG. 10(d), FIG. 10(f) and FIG. 10(h) are schematic diagrams of a second design. FIG. 10(b) shows three beams (each beam is indicated by dotted-line boundaries and an arc) emitted from one point which can pass through a collimator. FIG. 10(d) shows first parallel beams which can pass through the collimator (emitting towards lower right). FIG. 10(f) shows second parallel beams which can pass through the collimator (a direction parallel to the holes). FIG. 10 (h) shows third parallel beams which can pass through the collimator (emitting towards upper right). In the two designs, a total thickness of the collimator is the same, the collimation layers have the same thickness, and adjacent layers are staggered. Because of different thicknesses of each collimation layer and different number of layers, an inclination of the first parallel beams and the third parallel beams relative to the direction of holes in the second design is greater than that in the first design, and the beam is thinner in the second design, leading to a better resolution. As shown in FIG. 11, twelve collimation layers are included in this design. In a case that the twelve collimation layers are aligned, as shown in FIG. 11(a), the collimator is equivalent to a single layer design collimator having the same thickness. The results shown in FIG. 11(b), FIG. 11(c) and FIG. 11(d) are obtained by changing the staggering manner. FIG. 11 (b) shows a case that all adjacent layers are staggered. FIG. 11(c) is equivalent to the design in FIG. 10(a). FIG. 11(d) shows a more complex adjustment result, in which a staggering amount of adjacent layers is 0 or ⅓ of a pitch of holes. Because of the different staggering, the different designs have different collimation performances to be adapted to different applications. FIG. 11(a) only shows one state of the given twelve-layer adjustable collimator. It can be seen from FIG. 7, FIG. 9 to FIG. 11 that, multiple collimation layers may be staggered in multiple manners. For example, in the staggered multiple collimation layers, odd numbered collimation layers are aligned to each other, and even numbered collimation layers are aligned to each other. For another example, the staggering manner of the multiple collimation layers may be that all adjacent collimation layers are staggered with each other. Alternatively, the multiple collimation layers are grouped into multiple groups in order, and the multiple groups are staggered with each other. The number of collimation layers included in each group may be the same or different. FIG. 11(b) shows a case that adjacent collimation layers are staggered. In FIG. 11(c), the collimation layers are grouped into six groups, each group includes two collimation layers, and the six groups are staggered with each other. FIG. 11(d) shows a relatively complex case of eight groups in total, in which a single collimation layer may also be counted as a group. From left to right, both the third group and the fifth group include two collimation layers, and the eighth group includes three collimation layers. The figures show multiple cases, which are not described respectively herein. The collimator according to the present disclosure includes Q collimation layers having adjustable coupling relationship, and a value of Q ranges from 2 to 30. Optional collimation layer types may include a type of parallel hole collimator, a type of convergent collimator and a type of divergent collimator. Collimation holes may have a shape of regular triangle, square, regular hexagon, round, polygon approximate to a round, and the like. Based on the type of holes on the collimation layer and the staggering design mentioned above, the collimator may be adjusted into multi-layer staggered collimators with different structures, thereby obtaining collimation results of different performances. In addition, a part of the adjustable collimation layers may be aligned to combine adjacent layers into one layer, or all of the adjustable collimation layers may be aligned to combine the Q layers into one layer. Compared with the single layer collimator, the multi-layer staggered coupling collimator according to the present disclosure not only has improved performances and multiple selectable performances, but also has good machinability. Since the collimator includes multiple collimation layers, each of the multiple collimation layers is thin, which ensures machining precision.
042004926	abstract	A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.
	claims	1. An accelerator neutron source (ANS), the ANS comprising:a cylindrical housing comprising a cavity containing a gas comprising deuterium and tritium;a field ionization (FI array) within the cavity and configured to generate deuterium and tritium ions from the gas;a target comprising implanted ions within the cavity, the target configured to receive the deuterium and tritium ions and generate neutrons; anda cylindrical grid accelerating electrode within the cavity, the accelerating electrode coaxial with the housing and configured to direct the deuterium and tritium ions to a central region of the cavity within the cylinder of the accelerating electrode and aligned with a central axis of the housing and the accelerating electrode and to eject the ions along the central axis toward the target. 2. The ANS of claim 1, wherein the FI array further comprises plurality of nanotips positioned about a substrate, the nanotips extending from an inner surface of the housing into the cavity. 3. The ANS of claim 2, wherein the nanotips of the FI array are positioned cylindrically about the inner surface of the housing. 4. The ANS of claim 1, wherein the FI array, the accelerating electrode, and the target are operably connected and configured to produce about 10{circumflex over ( )}9 neutrons per second. 5. The ANS of claim 1, wherein the FI array further comprises an ionization voltage source configured to apply a voltage to the electrodes of the FI array. 6. The ANS of claim 1, further comprising an extraction ring configured to extract deuterium and tritium ions from the central region of the cavity within the cylinder of the accelerating electrode. 7. The ANS of claim 1, further comprising an ultra-high voltage source of about 110 kV connected to the target and configured to accelerate the deuterium and tritium ions from the central region of the cavity within the cylinder of the accelerating electrode toward the target.
042474957	claims	1. Method for the manufacture of UO.sub.2 nuclear fuel pellets containing PuO.sub.2 in set amounts, which pellets are soluble in nitric acid, which comprises (a) mixing uranium oxide powder having oxygen in stoichiometric excess of the dioxide, with plutonium dioxide powder in an amount of 15 to 50% plutonium dioxide by weight of the mixture of uranium oxide and plutonium dioxide, (b) milling the mixture of uranium oxide powder and plutonium dioxide powder and pressing the milled mixture to form pellets or granules (c) sintering the pellets in a reducing atmosphere in a furnace, comminuting the sintered pellets to primary grain sizes of less than 2 .mu.m by milling, pressing the comminuted grains to form pellets, and comminuting the pellets to free-flowing granules, (d) mixing the free-flowing granules with uranium oxide granules in an amount to obtain a desired UO.sub.2 /PuO.sub.2 ratio in the resultant mixture, and (e) pressing the resultant mixture into pellets and sintering the pellets to form UO.sub.2 nuclear fuel pellets containing PuO.sub.2 soluble in nitric acid. (a) mixing uranium oxide powder having oxygen in stoichiometric excess of the dioxide, with plutonium dioxide powder in an amount of 15 to 50% by weight of the mixture of uranium oxide and plutonium dioxide to obtain a desired UO.sub.2 /PuO.sub.2 ratio in the resultant mixture suitable for producing UO.sub.2 nuclear fuel pellets. containing PuO.sub.2 in set amounts for fast nuclear reactors, (b) milling the mixture of uranium oxide powder and plutonium dioxide powder and pressing the milled mixture to form pellets, (c) sintering the pellets in a reducing atmosphere in a furnace, comminuting the sintered pellets to primary grain sizes of less than 2 .mu.m by milling, pressing the comminuted grains to form pellets and comminuting the pellets to free-flowing granules, (d) and pressing the free-flowing granules into pellets and sintering the pellets to form UO.sub.2 nuclear fuel pellets containing PuO.sub.2 soluble in nitric acid and suitable for fast nuclear reactors. 2. Method according to claim 1, wherein the process product of step (c), the free-flowing granules is at least in part sent to storage and is stored until processed further into nuclear fuel pellets of desired, possibly different plutonium content. 3. Method according to claim 1 or claim 2, wherein the UO.sub.2 granules to be added according to step (d) is adjusted as to density, grain shape and sinterability in accordance with the UO.sub.2 /PuO.sub.2 free-flowing granules of step (c). 4. Method for the manufacture of UO.sub.2 nuclear fuel pellets containing PuO.sub.2 in set amounts suitable for fast nuclear reactors, which pellets are soluble in nitric acid, which comprises,
043550029	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS The present invention has been attained by examining the possibility of failure of nuclear fuel element containing burnable poison and of nuclear fuel element containing no burnable poison. Failure of nuclear fuel elements will be explained below. When nuclear fuel assembly is loaded within core of a nuclear reactor and operation of the nuclear reactor is started, there is the possibility of the nuclear fuel elements being deformed and failed as mentioned below. That is, deformation of columnar UO.sub.2 pellets 3 charged in fuel cladding 2 of nuclear fuel element 1 acts on the fuel cladding 2 to cause deformation of the fuel cladding 2 as shown in FIG. 1. UO.sub.2 pellets 3 are deformed due to the difference in thermal expansion caused by the difference in temperature, i.e., temperature of the central part of UO.sub.2 pellets 3 being higher than that of the circumferential part and due to expansion of the volume of UO.sub.2 pellets 3 caused by accumulation of fission products in the pellets. Thus, the cross sectional area of UO.sub.2 pellets 3 at both ends becomes greater than that of UO.sub.2 pellets 3 at the central portion in the axial direction. Ridge portion 4 is formed at the outer circumference of both ends of UO.sub.2 pellets and thus the cross-sectional area at said portions of UO.sub.2 pellets 3 is increased. In this way, deformation of UO.sub.2 pellets 3 proceeds and UO.sub.2 pellets 3 are allowed to contact with fuel cladding 2. Soon, fuel cladding 2 is expanded out by the ridge portions 4 formed at UO.sub.2 pellets 3. Thus, ridge portions 5 are formed at fuel cladding 2, too and appearance of fuel cladding 2 becomes like that of a bamboo, namely, the fuel cladding appears to have nodes. Rate of thermal expansion of UO.sub.2 pellets 3 is greater than that of fuel cladding 2. Therefore, when the deformation as mentioned above occurs in fuel cladding 2, a tensile stress occurs in fuel cladding 2 in its axial direction due to thermal expansion of UO.sub.2 pellets 3. Furthermore, owing to the difference in thermal expansion of central part and circumferential part of UO.sub.2 pellets 3, a compressive stress occurs at the central part of a high temperature and a tensile stress at the circumferential part of a low temperature. Therefore, as shown in FIG. 2, cracks 6 are formed in UO.sub.2 pellets 3 and a tensile stress occurs also in the circumferential direction of fuel cladding 2. Interaction of the tensile stresses in axial direction and in circumferential direction which occur in fuel cladding 2 causes concentration of the greatest local strain at the portions of ridge portions 5 of fuel cladding 2 which face cracks 6. In this way, failure of fuel cladding 2 occurs. In the nuclear fuel assembly loaded in a nuclear reactor are charged gadolinium-nuclear fuel elements containing burnable poison gadolinium oxide and nongadolinium-nuclear fuel elements containing no burnable poison. The nongadolinium-nuclear fuel element is arranged adjacent to the gadolinium-nuclear fuel element. After said nuclear fuel assembly was loaded in a core of a nuclear reactor, the nuclear reactor was operated and amount of elongation of the fuel cladding of each nuclear fuel element was measured. Uranium-235 of about 2.8% in enrichment and 1.5% of gadolinium oxide were contained in the gadolinium-nuclear fuel element and uranium-235 of about 2.8% in enrichment was contained in the nongadolinium-nuclear fuel element which contained no gadolinium oxide. Said uranium-235 is a fissionable material. FIG. 3 shows the relation between burn-up and elongation of the fuel cladding in its axial direction (which is referred to as merely elongation hereinafter) in the gadolinium-nuclear fuel element and nongadolinium-nuclear fuel element. Curve 8 indicates the changes in elongation of the fuel cladding of the gadolinium-nuclear fuel element. Curve 10 indicates the changes in elongation of the fuel cladding of the nongadolinium-nuclear fuel element. The elongation of the gadolinium-nuclear fuel element is smaller than that of the nongadolinium-nuclear fuel element at a low burn-up. However, when the burn-up exceeds about 3000 megawatt.day/ton, the elongation of the gadolinium-nuclear fuel element becomes larger than that of the nongadolinium-nuclear fuel element. The amount of gadolinium oxide in the gadolinium-nuclear fuel element decreases with increase in the burn-up. The elongation of the fuel cladding of the gadolinium-nuclear fuel element increases with decrease in the amount of gadolinium oxide. The elongation of the fuel cladding of the gadolinium-nuclear fuel element reaches maximum value, namely, about 0.0612% at a burn-up at which gadolinium oxide is lost, namely, a burn-up of about 5000-6000 megawatt.day/ton. After the elongation of the fuel cladding of the gadolinium-nuclear fuel element reaches the maximum point A, the elongation decreases with increase in the burn-up and closes to the elongation of the fuel cladding of the nongadolinium-nuclear fuel element. This is due to the stress relaxing phenomenon caused by creep deformation of the gadolinium-nuclear fuel element. The elongation of the fuel cladding of the nongadolinium-nuclear fuel element is about 0.0276% at the burn-up at which the elongation of the fuel cladding of the gadolinium-nuclear fuel element reaches the maximum point A. That is, the elongation of the fuel cladding of the gadolinium-nuclear fuel element is markedly greater than that of the fuel cladding of the nongadolinium-nuclear fuel element at a burn-up of about 5000-6000 megawatt.day/ton. The elongation of the fuel cladding occurs due to both thermal expansion and elongation of UO.sub.2 pellets in the axial direction as the result of action of deformation of UO.sub.2 pellets on the fuel cladding. Elongation of the fuel cladding of the gadolinium-nuclear fuel element occurs mainly due to the latter cause. When the elongation due to the latter cause is great, tensile stress which occurs at ridge portions 5 of fuel cladding 2 is increased to result in increase in the possibility of failure of the nuclear fuel element. Thus, the possibility of the gadolinium-nuclear fuel element being failed is greater than the possibility of the nongadolinium-nuclear fuel element being failed. As the result of various researches for reducing the possibility of the gadolinium-nuclear fuel element being failed, it has been found that reduction of the failure can be attained by rendering the content of fissionable material in the gadolinium-nuclear fuel element smaller than that of fissionable material in the nongadolinium-nuclear fuel element. One suitable example of the present invention will be explained with reference to FIG. 4. FIG. 4 shows cross-section of nuclear fuel assembly 11. Nuclear fuel assembly 11 comprises channel box 13 in which many nuclear fuel elements are loaded. Although not shown in the drawing, lower part of nuclear fuel elements 15 is attached to a lower tie plate and upper part thereof is attached to an upper tie plate. Nuclear fuel element 15 is held by spacers between the upper and lower parts. Cooling water flows through channel box 13 of the nuclear fuel assembly arranged in the core of the nuclear reactor. Nuclear fuel elements in nuclear fuel assembly 11 are roughly classified into gadolinium-nuclear fuel elements 16 and nongadolinium-nuclear fuel elements 17 as mentioned hereinbefore. In this example, nongadolinium-nuclear fuel elements 17 are also classified into four kinds depending upon the difference in enrichment of uranium-235 which is a fissionable material contained in the elements. That is, in nuclear fuel assembly 11 there are loaded nongadolinium-nuclear fuel elements 17A containing uranium-235 of about 2.8% in enrichment, nongadolinium-nuclear fuel element 17B containing uranium-235 of about 2.1 % in enrichment, nongadolinium-nuclear fuel elements 17C containing uranium-235 of about 1.8% in enrichment and nongadolinium-nuclear fuel elements 17D containing uranium-235 of about 1.4% in enrichment. Many nuclear fuel assemblies 11 are regularly arranged in lattice state in the core of a nuclear reactor. Control rod 19 is inserted in the spaces formed by the four nuclear fuel assemblies 11 for controlling the power of the nuclear reactor. Near side walls 21 and 22 of channel box 13 which are opposite to control rod 19, mainly nongadolinium-nuclear fuel elements 17B, 17C and 17D are disposed. Nongadolinium-nuclear fuel elements 17A are disposed in the remaining area of channel box 13 and up to near side walls 23 and 24 of channel box 13 which are opposite to side walls 21 and 22. Such disposition of nongadolinium-nuclear fuel elements 17 results from consideration on smoothing the power of nuclear fuel assembly 11 in horizontal direction. Gadolinium-nuclear fuel elements 16 contain about 1.5% of gadolinium oxide and moreover contain uranium-235 of about 2.0% in enrichment. The three gadolinium-nuclear fuel elements 16 are disposed between nongadolinium-nuclear fuel elements 17 as shown in FIG. 4. Disposition of gadolinium-nuclear fuel elements 16 and nongadolinium-nuclear fuel elements 17 in channel box 13 is the same as the conventional manner. Furthermore, enrichment of uranium-235 contained in each nongadolinium-nuclear fuel element 17 is also the same as in the conventional elements. Enrichment of uranium-235 contained in gadolinium-nuclear fuel element 16 is lower than that of uranium-235 contained in non gadolinium-nuclear fuel element 17A which is adjacent to the gadolinium-nuclear fuel element 16. Such reduction in enrichment means reduction in content of uranium-235 in the nuclear fuel elements. When a nuclear reactor having nuclear fuel assembly 11 disposed in a core is operated, elongations of fuel claddings of gadolinium-nuclear fuel element 16 and nongadolinium-nuclear fuel element 17A adjacent to the element 16 are as shown in FIG. 3. Elongation of the fuel cladding of nongadolinium-nuclear fuel element 17A increases with increase in the burn-up as shown by curve 10 as in the conventional manner. On the other hand, elongation of the fuel cladding of gadolinium-nuclear fuel element 16 changes with increase in the burn-up as shown by curve 26. Elongation of the fuel cladding of gadolinium-nuclear fuel element 16 also becomes maximum at a burn-up of about 5000-6000 megawatt.day/ton like elongation of a fuel cladding of conventional gadolinium-nuclear fuel element which is shown by curve 8. However, the value of the maximum point B of curve 26 is markedly smaller than the value of maximum point A. Moreover, at the same burn-up, elongation of the fuel cladding at the maximum point B of curve 26 is smaller than that of the fuel cladding of nongadolinium-nuclear fuel element 17A adjacent to gadolinium-nuclear fuel element 16. This is because enrichment of uranium-235 contained in gadolinium-nuclear fuel element 16 is about 2.0%, which is lower than about 2.8% of uranium-235 contained in nongadolinium-nuclear fuel element 17A adjacent to element 16. Elongation of the fuel cladding of gadolinium-nuclear fuel element 16 becomes closest to that of the fuel cladding of nongadolinium-nuclear fuel element 17A at a burn-up of about 5000-6000 megawatt.day/ton, but the former is extremely lower than the latter at a burn-up outside said range. Therefore, the possibility of gadolinium-nuclear fuel element 16 being failed is markedly reduced. Furthermore, the number of times for substitution of nuclear fuel assembly 11 disposed in a core with fresh nuclear fuel assembly is decreased and the possibility of reduction in operation rate of nuclear reactor is decreased. Enrichment of uranium-235 contained in gadolinium-nuclear fuel element 16 should preferably be less than about 72% of enrichment of uranium-235 contained in nongadolinium-nuclear fuel element 17A adjacent to element 16. Thus, the possibility of gadolinium-nuclear fuel element 16 being failed can be reduced to less than the possibility of nongadolinium-nuclear fuel element 17A being failed by previously adjusting enrichment of uranium-235 contained in gadolinium-nuclear fuel element 16. When enrichment of uranium-235 contained in gadolinium-nuclear fuel element 16 is about 2.4%, elongation of the fuel cladding of gadolinium-nuclear fuel element 16 at the maximum point B is greater than that of the fuel cladding of nongadolinium-nuclear fuel element 17A adjacent to element 16. However, when enrichment of uranium-235 contained in gadolinium-nuclear fuel element 16 is less than about 2.8%, the value of the maximum point B is smaller than that of the maximum point A. That is, when enrichment of uranium-235 contained in gadolinium-nuclear fuel element 16 is less than that of uranium-235 contained in nongadolinium-nuclear fuel element 17A adjacent to element 16, the possibility of gadolinium-nuclear fuel element 16 being failed becomes smaller than the conventional nuclear fuel element. As nuclear fuel assembly 11, the possibility of failure of this nuclear fuel assembly 11 is also reduced than the conventional assembly. As mentioned before, content of uranium-235 in the nuclear fuel element can be made lower by decreasing enrichment of fissionable material contained in the nuclear fuel elements, namely, uranium-235. Furthermore, content of uranium-235 in the nuclear fuel element can be made small by decreasing packing density of uranium dioxide in the nuclear fuel element. FIG. 5 shows gadolinium-nuclear fuel element 30 which is applied to a nuclear fuel assembly and which is another example of the present invention. Gadolinium-nuclear fuel element 30 comprises fuel cladding 32, UO.sub.2 pellets 34 and 35 and end plugs 37 and 38. UO.sub.2 pellets 34 and 35 are loaded in fuel cladding 32. End plugs 37 and 38 are attached to both ends of fuel cladding 32 and are sealed by welding. Coil spring 40 is disposed at the upper part of gadolinium-nuclear fuel element 30. Coil spring 40 holds down UO.sub.2 pellets 34 and 35 through wafer 41. Enrichment of uranium-235 contained in UO.sub.2 pellets 34 is about 2.0%. Enrichment of uranium-235 contained in UO.sub.2 pellets 35 is about 2.8%. About 1.5% of gadolinium oxide is contained in UO.sub.2 pellets 34 and 35. UO.sub.2 pellets 34 are disposed in the central part of gadolinium-nuclear fuel element 30. UO.sub.2 pellets 35 are disposed at the upper and lower parts of gadolinium-nuclear fuel element 30. Gadolinium-nuclear fuel elements 30 are inserted in place of above mentioned gadolinium-nuclear fuel elements 16 to constitute a nuclear fuel assembly. Other constitutions of the nuclear fuel assembly are the same as those of nuclear fuel assembly 11. In the upper and lower parts of gadolinium-nuclear fuel element 30, there is contained uranium-235 having the same enrichment as that of uranium-235 contained in nongadolinium-nuclear fuel element 17A which is adjacent to element 30. However, enrichment of uranium-235 contained in the central part of gadolinium-nuclear fuel element 30 is lower than that of uranium-235 contained in nongadolinium-nuclear fuel element 17 which is adjacent to element 30. When the nuclear fuel assemblies are loaded in core of a nuclear reactor and operation of the nuclear reactor is started, distribution of power of the nuclear fuel element in its axial direction is high in the central part of the nuclear fuel element and low in the upper and lower parts. Therefore, the possibility of failure in the upper and lower parts of the nuclear fuel element is low. This example which was made taking the above matters into consideration results in the similar effects to those of the example shown in FIG. 4. Furthermore, in this example, since enrichment of uranium-235 present in both end parts of gadolinium-nuclear fuel element 30 is high, there is obtained the effect that excess reactivity becomes greater than in the previous example. The present invention can also be applied when boron, cadmium, erbium, europium, hofnium, samarium, chemical compounds of them, and other non-fissionable materials having high thermal neutron absorption cross section are used in place of gadolinium oxide as a burnable poison. The present invention can also be applied to nuclear fuel assemblies of nuclear reactors other than boiling water reactors. According to the present invention, the possibility of failure of nuclear fuel element containing burnable poison can be reduced and reliability of nuclear fuel assembly can be improved. This also results in increase in safety of nuclear reactors.
	summary
	description	This application claims priority to and the benefit of Korean Patent Application No. 10-2013-0084657 filed in the Korean Intellectual Property Office on Jul. 18, 2013, the entire contents of which are incorporated herein by reference. (a) Field of the Invention The present invention relates to a decay heat removal system for cooling a nuclear power plant, and more particularly to a decay heat removal system with a hybrid heat pipe having a neutron absorber and a coolant for cooling a reactor core, which is installed in a nuclear reactor vessel, a nuclear fuel storage facility or the like nuclear power plant, and the hybrid heat pipe having the neutron absorber and the coolant is used to remove decay heat of a reactor core arranged in a nuclear reactor vessel or remove decay heat of spent fuel. (b) Description of the Related Art In general, when an accident occurs in a nuclear reactor, a coolant is injected to a primary system and cools a heated nuclear reactor vessel in order to remove decay heat generated from a reactor core arranged in the reactor vessel. However, if the supply of the coolant is restricted by the accident, it is impossible to remove the decay heat of the reactor core. Further, if a corium leaks out of the nuclear reactor vessel even though the coolant is supplied, a problem can arise in that a secondary accident such as a steam explosion may occur due to vapor from evaporation of the coolant inside pressure vessel. Also, the decay heat of the reactor core is removed by an indirect cooling method that the coolant supplied through the primary system cools the nuclear reactor vessel while being in contact with an external surface of the nuclear reactor vessel, and therefore a problem arises in that an efficiency of removing the decay heat of the reactor core is decreased. Patent Document 1. Korean Patent Publication No. 2013-0047871 (May 9, 2013), entitled ‘device for residual heat removal of integrated reactor and its method.’ Accordingly, the present invention is conceived to solve the forgoing problems, and an aspect of the present invention is to provide a decay heat removal system with a hybrid heat pipe having a neutron absorber and a coolant for cooling a nuclear power plant, in which occurrence of a secondary accident is prevented due to vapor because decay heat of a reactor core arranged in a reactor vessel is removed by the heat pipe, and the decay heat is removed through direct contact with the reactor core or nuclear fuel generating the decay heat. In accordance with an aspect of the present invention, there is a decay heat removal system to remove decay heat of a reactor core arranged in a nuclear reactor vessel, the decay heat removal system including: a first heat pipe which is placed in an upper plenum of the reactor vessel and arranged in upward and downward directions corresponding to a position of an insertion hole formed on a top of the reactor vessel; a control rod drive mechanism which is connected to an upper plenum of the first heat pipe and drives the first heat pipe to move up and down so that the first heat pipe can be selectively inserted in a control rod insertion hole of the reactor core arranged in the nuclear reactor vessel; and a second heat pipe which is coupled to and in close contact with a bottom surface of the nuclear reactor vessel and removes the decay heat generated in the reactor core. Here, the first heat pipe may have a heat sink with the coolant from an upper plenum of the nuclear reactor vessel as a condenser, or be connected to an independent condenser cooling tank, and perform cooling by absorbing the decay heat generated in the reactor core, and transferring the absorbed heat to the coolant Here, the temperature of the coolant in an upper plenum, i.e., in a condenser of the first heat pipe may be maintained or adjusted by convection of the coolant based on connection between the upper plenum and an in-containment refueling water storage tank, cooling based on a heat exchanger using heat pipes additionally installed in the upper plenum, or cooling based on a cooling tank provided on an outer wall of the upper plenum. Here, the second heat pipe may include a first end coupled to a storage tank provided in a containment building, and a second end attached to the bottom surface of the nuclear reactor vessel, and perform the cooling by receiving the coolant stored in the storage tank to the condenser, absorbing the decay heat generated in the reactor core and transferring the absorbed heat to the coolant. Also, the second heat pipe may be made of a flexible material, be curved corresponding to the shape of the bottom surface of the nuclear reactor vessel, and closely contact the bottom surface. Also, the second heat pipe may have a bellows-like structure, be curved corresponding to the shape of the bottom surface of the nuclear reactor vessel, and closely contact the bottom surface. Also, a working fluid circulated in the first heat pipe or the second heat pipe may include one of water (H2O), a nanofluid, a refrigerant, mercury (Hg), lithium (Li) and FLiBe (LiF-BeF2). Also, a wick placed in the first heat pipe or the second heat pipe may include one of carbon fiber, copper, stainless steel, zirconium alloy, silicon carbide (SiC), and boron carbide (B4C). Also, a case material forming an outer appearance of the first heat pipe or the second heat pipe includes one of stainless steel, zirconium alloy, inconel alloy and molybdenum alloy. In accordance with an aspect of the present invention, there is provided a decay heat removal system to remove decay heat of a spent nuclear fuel, the decay heat removal system including: a plurality of unit storage racks which includes a plurality of standing heat pipe plates coupled to one another and assembled in the form of a box to internally form a spent fuel storage for storing the nuclear fuel, in which a working fluid is circulated inside the heat pipe plate and absorbs the decay heat of the spent fuel. Here, the respective unit storage racks may be arranged and close to each other in the storage space where a coolant for removing the decay heat of the spent fuel. Also, the decay heat removal system may further include a third heat pipe inserted in a control rod insertion hole penetrating a middle of the fuel assemblies, and internally circulating a working fluid to absorb the decay heat from the inside of the nuclear fuel. In accordance with an aspect of the present invention, there is provided a decay heat removal system for cooling the nuclear power plant to remove decay heat of spent nuclear fuel, the decay heat removal system including: a storage container which forms an internal space for storing the spent fuel assemblies; a fourth heat pipe which is horizontally arranged in the form of traversing the internal space of the storage container, and circulates a working fluid therein to absorb the decay heat of the spent fuel; and a fifth heat pipe which is horizontally arranged in the form of traversing the internal space of the storage container, is installed in a direction of intersecting the fourth heat pipe, and, together with the fourth heat pipe, forms a rectangular storage space in which the spent fuel is arranged. Here, the fourth heat pipes and the fifth heat pipes may be spaced apart from each other at a distance corresponding to a horizontal length of a storage unit for the spent fuel, and closely contact circumference of the nuclear fuel. Also, the storage container may be supplied with and store a coolant for removing the decay heat of the spent fuel assemblies. Meanwhile, the decay heat removal system may further include a sixth heat pipe inserted in a control rod insertion hole penetrating the middle of the nuclear fuel assembly, and internally circulating a working fluid to absorb the decay heat from the inside of the spent fuel. Hereinafter, exemplary embodiments according to the present invention will be described with reference to accompanying drawings. Also, terms and words used in the following description and claims have to be interpreted by not the limited meaning of the typical or dictionary definition, but the meaning and concept corresponding to the technical idea of the present invention on the assumption that the inventor can properly define the concept of the terms in order to describe his/her own invention in the best way. First, a decay heat removal system with a hybrid heat pipe having a neutron absorber and a coolant for cooling reactor core (hereinafter, referred to as a ‘decay heat removal system’) according to a first embodiment of the present invention will be described with reference to FIG. 1. According to the first exemplary embodiment, the decay heat removal system for cooling the nuclear power plant is installed in a nuclear reactor vessel 10, and removes decay heat of a reactor core 11 arranged in the reactor vessel 10. As shown in FIG. 1, the decay heat removal system for cooling a reactor core includes a first heat pipe 110, a control rod drive mechanism 120 and a second heat pipe 130. The first heat pipe 110 is a cooling means inserted in the reactor core 11 arranged inside the reactor vessel 10 and directly removing the decay heat generated from the reactor core 11. Further, the first heat pipe 110 is placed above the reactor vessel 10 and arranged in upward and downward directions so as to correspond to a position of an insertion hole formed on a top of the reactor core 11. Here, the first heat pipe 110 is inserted into the reactor core 11 through the insertion hole formed on the top of the reactor vessel 10, and has a lower portion to be inserted in a control rod insertion hole 12 formed in the reactor core 11 arranged inside the nuclear reactor vessel 10, thereby removing the decay heat of the reactor core 11. To this end, the first heat pipe 110 has an outer diameter corresponding to a control rod to be inserted in the reactor core 11, and the spaced positions among the first heat pipes 110 correspond to spaced positions among a plurality of control rod insertion holes 12 formed in the reactor core 11 so that the first heat pipes 110 can be respectively inserted in the control rod insertion holes 12, thereby removing the decay heat of the reactor core 11. Here, the control rod insertion hole 12 refers to a previously formed insertion hole in which the control rod to be arranged in the reactor vessel 10 can be inserted. Thus, there is no need of separately forming an additional insertion hole through which the first heat pipe 110 can be inserted in the reactor core 11, and therefore the first heat pipe 110 can be applied as a cooling facility to the existing nuclear reactor system without changing the design of the reactor core 11. The control rod drive mechanism 120 is a driving means connected to the first heat pipe 110 and driving the first heat pipe 110 to move up and down. That is, the control rod drive mechanism 120 is connected to the top of the first heat pipe 110 and drives the first heat pipe 110 to move up and down so that the first heat pipe 110 can be selectively inserted in the control rod insertion hole 12 of the reactor core 11 arranged in the nuclear reactor vessel 10. Here, the control rod drive mechanism 120 may be controlled in accordance with a control signal from a control system of a nuclear power plant. When an accident occurs, the control system outputs a control signal to the control rod drive mechanism 120 in order to move the first heat pipe 110 down, and the control rod drive mechanism 120 moves down in response to a received control signal, thereby inserting the first heat pipe 110 in the reactor core 11. Here, as shown in FIG. 5 the first heat pipe 110 receives the coolant from the upper plenum 14 of the nuclear reactor vessel 10 to a condenser, and transfers the decay heat generated in the reactor core 11 to the coolant, thereby performing the cooling. Here, the temperature of the coolant in the upper plenum, i.e., in the condenser of the first heat pipe 110 may be maintained or adjusted by convection of the coolant based on connection between the upper plenum and an in-containment refueling water storage tank, cooling based on a heat pipe heat exchanger additionally installed in the upper plenum, or cooling based on a condenser cooling tank provided on an outer wall of the upper plenum. For reference, FIG. 5 is a perspective view of the decay heat removal system for cooling a reactor core according to the present invention, in an upper plenum of the reactor vessel. The second heat pipe 130 is a cooling means for secondarily removing the decay heat of the reactor core 11 together with the first heat pipe 110. As shown in FIG. 1, the second heat pipe 130 is provided to closely contact a bottom surface of the nuclear reactor vessel 10 and absorbs the decay heat generated in the reactor core 11, thereby performing the cooling. Here, the second heat pipe 130 has a first end coupled to an in-containment refueling water storage tank 30 provided, and a second end attached to the bottom surface 13 of the reactor vessel 10. Further, the second heat pipe 130 receives the coolant 31 stored in the storage tank 30 to the condenser, absorbs the decay heat generated in the reactor core 11 and transfers it to the coolant 31, thereby performing the cooling. Also, the second heat pipe 130 is curved for the close contact along the hemispherical shape of the bottom surface 13 of the reactor vessel 10. To this end, the second heat pipe 130 is made of a flexible material and curved corresponding to the shape of the bottom surface 13 of the nuclear reactor vessel 10, thereby closely contacting the bottom surface 13. Further, the second heat pipe 130 may have a bellows-like structure, and be connected to and be in close contact with the bottom surface 13 as being curved along the shape of the bottom surface 13 of the nuclear reactor vessel 10. Further, the second heat pipe 130 is provided to be curved to change its shape freely. Therefore, even though a wall of the containment building 20 or other obstacles are placed between the reactor vessel 10 and the in-containment refueling water storage tank 30, the second heat pipe 130 can be curved corresponding to the shape of the wall or obstacle, thereby having an effect on being directly installed without changing the facility of the current nuclear power plant. Also, a working fluid circulated in the first heat pipe 110 or the second heat pipe 130 may include one of water (H2O), a nanofluid, a refrigerant, mercury (Hg), lithium (Li) and FLiBe (LiF-BeF2). A wick placed in the first heat pipe 110 or the second heat pipe 130 may include one of carbon fiber, copper (Cu), stainless steel, zirconium alloy, silicon carbide (SiC), and boron carbide (B4C). Further, a case material forming an outer appearance of the first heat pipe 110 or the second heat pipe 130 may include one of stainless steel, zirconium alloy, inconel alloy and molybdenum alloy. Also, the first heat pipe 110 or the second heat pipe 130 are provided in the form of a hybrid heat pipe internally provided with both the neutron absorber and the coolant, thereby not only absorbing neutrons generated in the reactor core 11 of the reactor vessel 10 but also removing the generated decay heat. At this time, referring to FIG. 4, a fin 111 may be provided in the condenser of the first heat pipe 110 and secure a heat transfer area, which employs air, water, nanofluid, seawater, nitrogen or liquid metal as a coolant for exchanging heat with the first heat pipe. In the decay heat removal system for cooling reactor core with the foregoing elements and functions according to the first embodiment of the present invention, the decay heat of the reactor core 11 arranged in the nuclear reactor vessel 10 is removed by the heat pipes 110 and 120, thereby preventing occurrence of a secondary accident due to vapor, and maximizing a cooling efficiency because the decay heat is removed by direct contact with the reactor core 11 generating the decay heat. Next, elements and functions of a decay heat removal system 40 for cooling the spent fuel according to a second embodiment of the present invention will be described with reference to FIG. 2. Referring to FIG. 2, the decay heat removal system 40 for cooling the spent fuel assemblies according to the second embodiment of the present invention is a cooling system for removing the decay heat of the spent nuclear fuel assemblies 41. As shown in FIG. 2, the decay heat removal system 40 has a plurality of unit storage racks 200 that includes a plurality of standing heat pipe plates 210 coupled to one another and assembled in the form of a box to thereby internally form a nuclear fuel storage 220 for storing the spent nuclear fuel assemblies 41, in which the working fluid is circulated inside the heat pipe plate 210 and absorbs the decay heat of the spent nuclear fuel assemblies 41. Here, the respective unit storage racks 200 are arranged and close to each other in the storage space where the coolant for removing the decay heat of the spent nuclear fuel assemblies 41 is stored. Therefore, the whole lateral sides of the spent nuclear fuel 41 inserted and stored in each unit storage rack are in direct contact with the inner surfaces of the heat pipe plate 210, thereby more quickly removing the decay heat. Also, the decay heat removal system 40 for cooling the spent fuel assemblies according to the second embodiment of the present invention may further include a third heat pipe 230 inserted in the control rod insertion hole 42 penetrating the middle of the spent fuel assemblies 41, and internally circulating the working fluid to absorb the decay heat from the inside of the spent fuel assemblies 41. Here, the control rod insertion hole 42 is a previously formed insertion hole in which the control rod can be inserted when the spent fuel assemblies 41 is arranged in the reactor vessel 10. Further, the heat pipe plate 210 and the third heat plate 230 are provided in the form of the hybrid heat pipe internally provided with both the neutron absorber and the coolant, thereby having functions of not only absorbing the neutron generated in the spent fuel assemblies 41, but also removing the generated decay heat. Also, a case material forming an outer appearance of the heat pipe plate 210 and the third heat plate 230 may include one of stainless steel, zirconium alloy, inconel alloy and molybdenum alloy, and a wick placed inside the heat pipe plate 210 and the third heat plate 230 and circulating the working fluid may include one of stainless steel, zirconium alloy, inconel alloy and molybdenum alloy. Also, the working fluid may include one of water (H2O), a nanofluid, a refrigerant, mercury (Hg), lithium (Li) and FLiBe (LiF-BeF2). In the decay heat removal system 40 for cooling spent fuel with the foregoing elements and functions according to the second embodiment of the present invention, it is possible to more quickly and stably remove the decay heat of the spent fuel assemblies 41 and to absorb and remove the unnecessarily generated neutrons as compared with the store rack made of a metal plate and storing the spent fuel assemblies 41. Next, elements and functions of a decay heat removal system 50 for cooling a nuclear power plant according to a third embodiment of the present invention will be described. Like the foregoing decay heat removal system 40 for cooling spent fuel assemblies according to the second embodiment, the decay heat removal system 50 for cooling the spent fuel assemblies according to the third embodiment of the present invention is a cooling system for removing the decay heat of the spent fuel assemblies 41. As shown in FIG. 3, the decay heat removal system 50 includes a storage container 51, a fourth heat pipe 310 and, a fifth heat pipe 320. The storage container 51 is a container internally formed with a storage space for storing the spent fuel assemblies 41 and placing each heat pipe 310, 320, 330 therein. Further, the coolant for removing the decay heat of the spent fuel assemblies 41 may be supplied to and stored in the storage container 51. The fourth heat pipe 310 is a cooling means arranged to be in close contact with each spent fuel assemblies 41 stored in the storage container 51 and removing the decay heat of the spent fuel assemblies 41. The fourth heat pipe 310 is horizontally arranged in the form of traversing the internal space of the storage container 51, and working fluid is circulated inside therein to absorb the decay heat of the spent fuel assemblies 41. Also, the fifth heat pipe 320 is a cooling means arranged to be in close contact with each spent fuel assemblies 41 stored in the storage container 51 together with the fourth heat pipe 310 and removing the decay heat of the spent fuel assemblies 41. The fifth heat pipe 320 is horizontally arranged in the form of traversing the storage space of the storage container 51. As shown in FIG. 3, the fifth heat pipe 320 is installed in a direction of intersecting the fourth heat pipe 310, and, together with the fourth heat pipe 310, forms a rectangular storage space in which the spent fuel assemblies 41 is arranged. Here, the fourth heat pipes 310 and the fifth heat pipes 320 are spaced apart from each other at a distance corresponding to a horizontal length of a storage unit for the spent fuel assemblies 41, and closely contact both sides of the spent fuel assemblies 41, thereby maximizing an efficiency of cooling the spent fuel assemblies 41. Also, the decay heat removal system 50 for spent fuel assemblies according to the third embodiment of the present invention may further include a sixth heat pipe 330 inserted in the control rod insertion hole 42 penetrating the middle of the spent fuel assemblies 41, and internally circulating the working fluid to absorb the decay heat from the inside of the spent fuel assemblies 41. Here, the control rod insertion hole 42 is a previously formed insertion hole in which the control rod can be inserted when the spent fuel assemblies 41 is arranged in the reactor vessel 10. Further, the fourth heat pipe 310, the fifth heat pipe 320 and the sixth heat pipe 330 are provided in the form of the hybrid heat pipe internally provided with both the neutron absorber and the coolant, thereby having functions of not only absorbing the neutron generated in the spent fuel assemblies 41, but also removing the generated decay heat. Also, the fourth heat pipe 310, the fifth heat pipe 320 and the sixth heat pipe 330 may use seawater as a final heat removing source, in which the seawater is circulated inside an independent condenser cooling tank 140 as shown in FIG. 4. For reference, FIG. 4 is a perspective view showing a configuration of a condenser cooling tank and a fin in the decay heat removal system for cooling the reactor core according to the present invention. Further, a case material forming an outer appearance of the fourth heat pipe 310, the fifth heat pipe 320 and the sixth heat pipe 330 may include one of stainless steel, zirconium alloy, inconel alloy and molybdenum alloy, and a wick placed inside the fourth heat pipe 310, the fifth heat pipe 320 and the sixth heat pipe 330 and circulating the working fluid may include one of stainless steel, zirconium alloy, inconel alloy and molybdenum alloy. Also, the working fluid may include one of water (H2O), a nanofluid, a refrigerant, mercury (Hg), lithium (Li) and FLiBe (LiF-BeF2). As described above, in the decay heat removal system with the hybrid heat pipe having the neutron absorber and the coolant for cooling reactor core and the spent fuel assemblies according to the present invention, the decay heat of the reactor core 11 arranged in the reactor vessel is removed the heat pipe, thereby preventing occurrence of a secondary accident due to vapor, and maximizing a cooling efficiency because the decay heat is removed by direct contact with the reactor core 11 generating the decay heat. Also, the second heat pipe placed beneath the reactor vessel is made of a flexible material or has a bellows-like structure, and thus curved along the shape of the bottom surface of the nuclear reactor vessel, thereby improving a cooling efficiency through close contact. Although a few exemplary embodiments of the present invention have been shown and described, it will be appreciated by those skilled in the art that changes may be made in these embodiments without departing from the principles and spirit of the invention, the scope of which is defined in the appended claims and their equivalents.
039403149	abstract	A nuclear reactor fuel element comprised of a plurality of fuel rods disposed in a plurality of spacers in which the tubular casing for each fuel rod is designed without regard to the mechanical stress produced by the spacers and has a reinforced wall thickness adjacent to the spacers which is thicker than the wall thickness of the tubular casing in other areas not adjacent to the spacers. The spacers are arranged in a circular mesh with a center support rod.
	summary
	summary
053965337	summary	BACKGROUND OF THE INVENTION The present invention is directed to a primary radiation diaphragm for a radiation apparatus. The diaphragm comprises at least one diaphragm plate adjustable in a plane perpendicular relative to the axis of the central ray of the radiation transmitter or emitter. U.S. Pat. No. 4,691,335, whose disclosure is incorporated herein by reference thereto and which claims priority from German Application 84 36 281(U), as well as German Published Application 14 41 312, disclose primary radiation diaphragms wherein two diaphragm plate pairs are respectively oppositely adjustable relative to one another in two parallel planes so that a rectangular gating of the x-ray beam is possible. The adjustment of the diaphragm plates lying above one another must, therefore, occur in such a way that only the opening that defines the radiation field is left free. The diaphragm plates are adjustably seated by, respectively, two mounts provided on sides lying opposite one another. SUMMARY OF THE INVENTION An object of the present invention is to provide a primary radiation diaphragm having adjustable diaphragm plates which is smaller, assembly-friendlier, more surveyable and has a less complicated structure. This object is inventively achieved in that a single mount is provided for the diaphragm plate and this mount engages the diaphragm plate and is adjustable along a guideway. An advantage of the invention is that the second mount of the diaphragm plate can, thus, be eliminated so that the requirements of the stated object are satisfied. When the ray beam of the radiation transmitter is to be gated slot-shaped, then it is advantageous when two diaphragm plates are arranged parallel to one another in the plane and are oppositely adjustable relative to one another. It is advantageous for precise guidance of the diaphragm plates when the guideway comprises two guide rails that are mounted on the same side of the housing of the primary radiation diaphragm and when the mount of each diaphragm plate respectively engages on the first and second guide rails via guide parts. When the radiation field is so gated quadratically or rectangularly, then it is advantageous when two additional diaphragm plates are provided in a plane parallel to the first diaphragm plates and offset by 90.degree. relative to the first diaphragm plates. These two additional diaphragm plates are also oppositely adjusted to one another. Other advantages and features of the invention will be readily apparent from the following description of the preferred embodiments, the drawings and claims.
	description	This invention was made with government support under DE-FOA-0000323 awarded by the US Department of Energy. The government has certain rights in the invention. -- The present invention relates to a system for generating isotopes useful for medical purposes, such as Mo-99, I-131, Xe-133, Y-90, Cs-137, I-125, and others, and in particular to a system employing a dry, granulated, uranium compound. Medical isotopes are employed in nuclear medicine where they may be administered to a patient in a form that localizes to specific organs or cellular receptors where they may be imaged with special equipment. Medical isotopes may also be used in the treatment of disease exploiting the tissue-destructive power of short-range ionizing radiation after such localization. Today, most radioisotopes used in nuclear medicine are produced in nuclear reactors employing highly enriched uranium (HEU). The reactors used for the production of Mo-99 for the United States are outside of the United States requiring the export of HEU and an attendant risk of nuclear proliferation associated with such out-of-country shipments. It has been proposed to generate medical isotopes using low enriched uranium (LEU) which cannot be used directly to manufacture nuclear weapons. Systems for this purpose are described in US patent applications: 2011/0096887 entitled: “Device and Method for Producing Medical Isotopes” and 2010/0284502 entitled: “High Energy Proton or Neutron Source” hereby incorporated by reference. In these systems, ions are directed through a target volume holding a gas to generate neutrons. The neutrons may expose a parent fissile material held in solution near the target volume in a fissile solution vessel. In one embodiment the target volume is annular and placed around a cylindrical fissile solution vessel holding the parent material solution. Ions are injected in a spiral through the target volume producing neutrons directed inwardly toward the parent material and outwardly toward a reflector. Neutrons received in the neutron rich parent fissile material (such as LEU uranium) experience a multiplication in which neutrons striking the parent material generate additional neutrons which strike additional neutron rich material in a chain reaction. In a nuclear reactor, at steady power, the effective neutron multiplication factor (keff) is equal to 1. In a subcritical system, keff is less than 1. One problem with aqueous reactors is that it can be difficult to maintain a stable power level. This is because there exist strong feedback mechanisms in the neutron multiplication factor as the temperature of the fissile solution rises and as voids are generated (gas bubbles caused by radiolysis breaking water into hydrogen and oxygen). The rapid reduction in the neutron multiplication factor results in a decrease in power, which causes the neutron multiplication factor to increase again. In particular, a control system that is trying to maintain constant power in the reactor may not be able to react sufficiently fast to adequately control the system. The result is a system with an unstable power level and potential safety impacts. Co-pending U.S. patent application Ser. No. 13/373,899 filed Dec. 5, 2011, hereby incorporated by reference in its entirety, describes an improved geometry for such aqueous reactors providing a fissile solution vessel in the form of a reduced thickness annulus. By controlling the aspect ratio of the annulus, improved reaction stability is employed and enhanced cooling provided. A significant disadvantage of these aqueous processes is that the fission fragments transfer substantial fission energy to the water medium. This energy causes the water molecules to break up into explosive hydrogen and oxygen gas and corrosive species such as hydrogen peroxide. The gases must be recombined in a separate system adding complexity to the aqueous processes and the concentration of hydrogen peroxide may build up in the water and may have to be controlled. Because the fission reactivity of these reactors decreases significantly as the water heats up and/or is radiolysed, and the density drops, recombined water must be returned to the vessel or new water added during operation. Also the pH of the solution has to be monitored and kept acidic to prevent the solution from precipitating. The present invention provides an improved method and apparatus for generating medical isotopes using a dry-phase granular uranium material such as a uranium compound such as uranium salt or uranium oxide. After irradiation in the dry state, the granular uranium material is dissolved in a solvent and the extraction of the medical isotopes may proceed as is done with aqueous reactors. Eliminating water from the granular uranium material during the irradiation reduces the risk of explosion from hydrogen and oxygen generated by radiolysis, the problems of pH control and water makeup, and other complexities attendant to aqueous reaction. The resulting process is more temperature stable and the processing can operate at temperatures higher than the boiling point of water for more efficient cooling. The granular uranium material is readily dissolved for simplified handling after irradiation. Specifically then, the present invention provides a method of producing medical isotopes by exposing a dry granular uranium material to radiation to produce the medical isotopes by nuclear reaction. The irradiated uranium material then dissolved in a solvent, typically an acid, and separated from the dissolved uranium material by standard aqueous separation techniques, to provide an isolated medical isotope. It is thus a feature of at least one embodiment of the invention to eliminate the disadvantages associated with aqueous solutions of uranium salt in producing hydrogen and oxygen through radiolysis such as creates: explosion risks, reaction stability problems, and the need for water makeup during production. It is further an object of the invention to eliminate the need for pH control of an aqueous solution and to avoid operating temperature limitations imposed in the processing of aqueous solutions. The method may include the step of recrystallizing the granular uranium material by removing the water and recycling the dry granular uranium material through the process again. It is thus a feature of at least one embodiment of the invention to provide efficient use of the uranium materials. The method may include the step of cooling the dry granular uranium material by fluid flow in thermal physical contact with the dry uranium salt during irradiation. It is thus a feature of at least one embodiment of the invention to provide a simple method of temperature control of the granular uranium material without the need for an aqueous solution. The granular uranium material may be a compound such as a uranium salt or uranium oxide. It is thus a feature of at least one embodiment of the invention to provide a system that may work with available and well-understood uranium compounds. The dry granular uranium material may be held in multiple containers and the solvent may be introduced into the containers to dissolve the reacted uranium granular uranium material in the containers for removal from the container in solution form. It is thus a feature of at least one embodiment of the invention to provide a simple method of transferring granular uranium material in dry form out of the containers. The method may include placement of one or more radiation reflectors near the dry granular uranium material during irradiation. It is thus a feature of at least one embodiment of the invention to enlist the radioactivity of the granular uranium material in supporting the desired reaction. The reaction vessel may include control elements that may be used to controllably absorb radiation and move the control elements to maintain at or near-critical reaction during irradiation. It is thus a feature of at least one embodiment of the invention to provide a system that may operate as a critical nuclear reactor greatly simplifying its construction. Alternatively, the radiation may be produced by an electrically powered neutron generator irradiating the dry granular uranium material in a sub-critical reactor. It is thus a feature of at least one embodiment of the invention to provide a system that may operate sub-critically for simple control. These particular objects and advantages may apply to only some embodiments falling within the claims and thus do not define the scope of the invention. Referring now to FIGS. 1a and 2, a medical isotope generator 10 per the present invention may provide for containers 12 into which may be placed a dry granular uranium material 14. As used herein, the term granular, refers to a collection of discrete macroscopic particles that may generally flow when poured for placement into the containers 12 without significant clumping caused by Van der Waals forces and that when held in the containers 12 preserve substantial airspace between particles in the filled container 12 that would permit infusion by a solvent. In one embodiment, the grains of dry uranium material 14 may be between 60 micrometers and two millimeters in dimension or alternatively between 125 micrometers and one millimeter in dimension. As depicted, the containers 12 may be cylindrical tubes closed at one end and having a lid 15 at the opposite end that may be used to provide a removable and replaceable watertight seal providing an enclosed volume within the container 12. The container 12 make be constructed of a radiation resistant alloy metal such as zircaloy-4 or the like. At this stage in the process, the granular uranium material 14 is substantially free of all liquid water. Preferably the granular uranium material 14 are loosely packed in the container 12, however, to be either readily removable by pouring or to allow for ready infusion of a solvent at a later stage of the process for dissolving the granular uranium material 14. The granular uranium materials may be a compound such as uranium oxide or uranium salt. The uranium salts may, for example, be uranyl nitrite or uranyl sulfate. Other salts such as uranyl fluoride and uranyl phosphate may also be used. The uranium oxides may, for example, be triuranium octoxide (U3O8), uranium dioxide (UO2), or uranium trioxide (UO3). Other uranium oxides may also be used. Referring now to FIGS. 1b and 2, the granular uranium material 14 as sealed in the container 12 may be irradiated by neutrons 16 to generate a desired medical isotope through a nuclear reaction caused by the impinging neutrons. The particular reaction may, for example, provide 99Mo isotope by fission of 235U in the granular uranium material 14. This process may be conducted at elevated temperatures exceeding the boiling point of water typically moderated by a cooling fluid flow, for example, around the outer surfaces of the container 12 or through channels formed inside the container 12 as will be discussed below. Referring to FIG. 1(c), after irradiation, the contained granular uranium material and medical isotope may be dissolved with water, an aqueous acid solution, or other solvent 17 to form a solution 18. In one embodiment, the solvent is nitric acid for uranyl nitrate and uranium oxides, and sulfuric acid is used for uranyl sulfate salts. For uranyl fluoride and uranyl phosphate the solvent could be hydrofluoric acid or phosphoric acid, respectively. The solution 18 may the product of a removal process in which the dry granular uranium material 14 are moved from the container 12 by flushing water through the interior volume of the container 12 into an accumulating container 20. Alternatively, the invention contemplates that the dry granular uranium material 14 may be removed from the container 12 by other means such as pouring or augering and then mixed with the solvent in the accumulating container 20. Referring now to FIG. 1(d), however formed, the solution 18 may then be provided to a separator 21 for separating out a medical isotope 22 from the water/uranium salt mixture 24. The separator 21, for example, may make use of adsorption column of titania or alumina. A LEU-modified Cintichem may then be used to purify the moly (primarily by iodine removal) after extraction from the solution to provide a source of the desired medical isotopes, particularly 99Mo. The medical isotope may be placed in containers 26 to be transported to a location of use. The remaining water/uranium salt mixture 24 may be recrystallized by a recycler 28, for example, by cooling the water/uranium salt mixture 24 to promote crystallization of the uranium salts followed by a decanting of the water and/or thermal evaporation of the water. The reconstituted dry uranium granular material 14 may then be replaced in a container 12 and this process repeated per the foregoing description. Referring now to FIG. 3, in one embodiment the containers 12 may be placed in a subcritical reactor 30, for example, providing a reaction chamber 31 having the dimensions of a cylindrical annulus and in which the containers 12 are placed in a circle at an equal distance from a central axis 29 of the reaction chamber 31 and oriented parallel to the central axis 29. The general construction of this subcritical reactor 30 will be similar to that described in the above co-pending U.S. patent application Ser. No. 13/373,899 with respect to features surrounding the reaction chamber 31. Specifically, electronically accelerated ions from an electronically controlled and powered ion accelerator 27 may pass downward through an axially extending trap assembly 32 to strike a target gas held within a cylindrical target chamber 33 concentric with the reaction chamber 31. The cylindrical target chamber 33 produces neutrons that pass radially outward into the reaction chamber 31 after passing through a coaxial and cylindrical wall 34 of a neutron multiplier/moderator material. The neutron multiplier/moderator material may be, for example, an aluminum-clad beryllium metal, or depleted uranium or other similar material that multiplies fast neutrons passing outward from the target chamber 33 and moderates fast neutrons traveling inward from the annular reaction chamber 31. The excess heat of the neutron multiplier/moderator 46 is removed by water jackets (not shown for clarity) which allow control of the temperature of the neutron multiplier/moderator 46 to ensure the escape of sufficient neutrons from the target chamber 33 while moderating neutrons received from the reaction chamber 31. In one embodiment, the neutron multiplier/moderator 46 may provide a 1.5-3.0 multiplication factor such as may be adjusted by adjusting its thickness. The outside of the reaction chamber 31 may be bounded by a cylindrical outer annular reflector 36 surrounding and coaxial with an annular reaction chamber 31. This reflector, for example, may be an aluminum walled chamber filled with a reflector material 38 which, in one embodiment, may be heavy water having a volume, for example, of 1000 liters. The reflector material 38 increases the generation efficiency by reflecting neutrons back into the reaction chamber 31 and therefore may also permit reaction control by draining water from the reflector 36 thus reducing the neutron reflection into the reaction chamber 31. This approach may also be used in controlling a critical reactor assembly as will be discussed below. Referring now to FIGS. 1(b) and 5, during operation, the irradiating neutrons 16 described above with respect to FIG. 1 will be produced both by the neutron ion accelerator 27 acting on the gas of the target chamber 33 and additional neutrons 16 generated in a chain reaction from the granular uranium material. Generally, the medical isotope generator 10 will be controlled to provide the desired level of neutron bombardment of the containers 12 necessary to produce the desired medical isotope per FIG. 1(b) as described above while operating with an effective neutron multiplication factor (keff) less than 1. This control may be affected by control of the ion accelerator 27 and/or control of moderating elements, for example, controlling the reflector 36 as described above with respect to water level or other control techniques described herein or well-known. During the irradiation process, the containers 12 and the contained dry granular uranium material may be controlled in temperature by natural convection or a pumping of a chilled fluid around them, for example, a chilled gas or liquids such as water. In the case where cooling water is used, a thin water jacket around each container 12 may be employed or the size of the reaction chamber 31 carefully limited to reduce the necessary volume of water and thereby suppress an undesirable feedback mechanism in neutron multiplication factor caused by voids generated in the water (gas bubbles caused by radiolysis breaking water into hydrogen and oxygen) changing the neutron multiplication factor. Referring now to FIGS. 1 and 4, in one embodiment, the containers may include an internal water circulation channel 40, for example a helix of metal tubing, that may receive and discharge cooling water 42 through external connections 44 (only one shown) to provide cooling of the contained granular uranium material 14 packed around the water circulation channel 40 with thermal connection thereto but in physical isolation. Rapid circulation, separation from the reactive materials and the limited volume of water minimizes the radiolysis problem. Other cooling liquids than water may be used including those providing cooling at higher operating temperatures than the boiling point of water. In addition or alternatively, each container 12 may provide for a pair of flushing connections 47, one that will receive water 42 from an external source and introduce it into the container 12, and one that will expel the received water into the accumulating container 20 during the extraction of the granular uranium material 14 from the container 12 per the discussion of FIG. 1(c). In this way, the removal of the granular uranium material 14 may be accomplished easily even if the granular uranium material 14 are packed or caked in the container 12 after irradiation. Generally the containers may remain in the reaction chamber 31 until needed, being subject to continual neutron bombardment. Referring also to FIG. 5, the above present invention may also work with a critical reactor 50 operating with an effective neutron multiplication factor (keff) equal to 1. In this system, multi-layered rows 52 of containers 12 may be placed in corresponding circles about a reflector core 54 and within a reflector shell 56 defining an annular space of a reaction chamber 58 similar to reaction chamber 31. The containers 12 may be a vertically oriented to be parallel to the axis 29 of the annular reaction chamber 58. One or both of the reflector core 54 and reflector shell 56 may include removable control rods 60 that may be used to control the resulting chain reaction and/or control rods (not shown) may be interspersed among the containers 12. In one embodiment, fuel rods 62 may also be interspersed in the rows 52 to promote the necessary nuclear reactions; however, it is believed that no such additional fuel rods 62 are required. The critical reactor 50 may otherwise be operated in a manner similar to a conventional nuclear reactor with the control rods 60 adjusted to promote the desired neutron multiplication factor using a feedback or other control system, for example, operating on an electronic computer according to a stored program. A simulation was performed on a double row of containers 12 arranged according to the parameters of following Table I in a critical reaction chamber per FIG. 4. TABLE IParameterQuantityUnitRow 1 Number of Tubes30Row 1 Tube Angular12.00degSpacingRow 1 Tube Axis Radial20.3160cmDistanceRow 1 Distance Between1.311cmTubesRow 2 Number of Tubes36Row 2 Tube Angular10.00degSpacingRow 2 Tube Axis Radial25.1320cmDistanceRow 2 Distance Between1.445cmTubesTube Inner Diameters2.54cmOxide Volume per Tube506.7ccTotal Oxide Volume33.443LitersTotal Uranium Mass207.519kgAdjusted Cold keff1.00000Adjusted Hot keff0.99957Fission rate1.496E+17f/sSpecific fission rate7.209E+14f/kgU-sIrradiation Time5.5days99Mo Activity EOI185329.3Ci As shown in FIG. 6, this configuration provides substantially higher specific fission rates, as indicated by curve 68, in comparison to a single row of containers 12 per curve 70 or a single row of larger diameter containers (1.25 inches) per curve 72. This is true over a range of different numbers of containers 12. The higher specific fission rate promotes increased efficiency in the generation of medical isotopes. With 66 containers 12, the medical isotope generator 10 could run continuously as a critical reactor 50 without the need of fuel rods 62 described above. It will be generally appreciated that the annular reaction chambers 31 and 58 need not be a cylindrical annulus but may take on other annular shapes such as a polygonal and that the term annular should be understood to include an annulus having an upper and/or lower solid base. Generally, each of the components including the neutron multipliers and reflectors may be cooled by water jackets that are not shown. The circulation of chilled water within the water jackets may be controlled by a feedback controller to control the temperature of the water to a predetermined value or to a dynamic value based on a monitoring of the general reaction rate by other means. In addition, the feedback controller may manage other control variables such as control of height of the water and/or control rods to control reaction rates. Generally, the medical isotope generator 10 will be further shielded with concrete and water according to standard practices. Other isotopes such as 131I, 133Xe, and 111In may also be produced by a similar structure. Referring now to FIG. 7, an alternative container 12 may provide for a lower discharge portion 70 positioned below a dump valve 72 separating the discharge portion 70 from an upper portion 71 normally holding the granular uranium material 14 and being, for example, a cylindrical tube into which the granular uranium material 14 is closely packed. In a closed state, the dump valve 72 holds the granular uranium material 14 in the upper portion 71. The dump valve 72 may be activated for example by an external guide linkage 74 or cable and may be biased by a spring or the like (not shown) to an open state for fail-resistant operation. In the open state, the dump valve 72 allows the granular uranium material 14 to drop rapidly into the lower portion 70 where the granular uranium material 14 is geometrically separated and/or isolated by neutron absorbing barriers 76 to rapidly quench an ongoing nuclear reaction. The system may be used independently or together with the control rods described above. Certain terminology is used herein for purposes of reference only, and thus is not intended to be limiting. For example, terms such as “upper”, “lower”, “above”, and “below” refer to directions in the drawings to which reference is made. Terms such as “front”, “back”, “rear”, “bottom” and “side”, describe the orientation of portions of the component within a consistent but arbitrary frame of reference which is made clear by reference to the text and the associated drawings describing the component under discussion. Such terminology may include the words specifically mentioned above, derivatives thereof, and words of similar import. Similarly, the terms “first”, “second” and other such numerical terms referring to structures do not imply a sequence or order unless clearly indicated by the context. When introducing elements or features of the present disclosure and the exemplary embodiments, the articles “a”, “an”, “the” and “said” are intended to mean that there are one or more of such elements or features. The terms “comprising”, “including” and “having” are intended to be inclusive and mean that there may be additional elements or features other than those specifically noted. It is further to be understood that the method steps, processes, and operations described herein are not to be construed as necessarily requiring their performance in the particular order discussed or illustrated, unless specifically identified as an order of performance. It is also to be understood that additional or alternative steps may be employed. It is specifically intended that the present invention not be limited to the embodiments and illustrations contained herein and the claims should be understood to include modified forms of those embodiments including portions of the embodiments and combinations of elements of different embodiments as come within the scope of the following claims. All of the publications described herein, including patents and non-patent publications are hereby incorporated herein by reference in their entireties.
046541845	summary	BACKGROUND OF THE INVENTION This invention pertains to methods and arrangements for attaining high beta values in plasma confinement devices. More specifically, this invention pertains to methods for accessing the second stability region of operation in toroidal magnetic confinement devices. The performance of a magnetic confinement device can be expressed by the parameter beta .beta., the ratio of the plasma kinetic pressure to the confining pressure of the magnetic field. Beta is a direct measure of the efficiency of the magnetic confinement; that is, high-.beta. systems make better use of the confining field than do low-.beta. systems. Beta is defined as: ##EQU1## where p.sub.av =.intg.pd.tau./.intg.d.tau. and B.sub.av.sup.2 =.intg.B.sup.2 d.tau./.intg.d.tau., the integration being over the plasma volume, where p.sub.av is the average plasma pressure and B.sub.av.sup.2 /2 is the average magnetic pressure. A plasma confined in a magnetic field may be unstable. Various instabilities have been predicted based on ideal single fluid magnetohydrodynamic (MHD) equilibrium and linear stability analyses in axisymmetric toroidal configurations. Potentially unstable MHD modes include: the ballooning modes, the Mercier modes (interchange modes), and external and internal kinks. Of these modes, ballooning and internal kinks are serious obstacles to creating and maintaining stable high-.beta. plasmas. Generally, the criteria for ideal MHD instability will depend on specific plasma parameters such as .beta., the pressure and safety factor profiles, and the various geometrical shaping factors. Consequently, stable operation has been limited to low betas. This region of operation is referred to as the "first region" of stable operation. Increasing beta beyond the limit of the first region results in operation in the unstable region where deleterious effects of unstable MHD modes are present. Several studies have been carried out to find environments favorable for suppressing the ballooning instability mode (e.g., A. M. Todd et al., Nucl. Fusion 19 743 (1979)). An empirical shape-optimization by Miller and Moore (Phys. Rev. Lett. 43, 765 (1979)) has shown that a strongly modified dee shaped plasma with an indentation on the inside edge of the plasma (i.e., inwardly concave at the inner-major-radius side) can enhance achievable stable .beta. against ballooning for small aspect ratio configurations. Similarly, Mercier (in Lectures in Plasma Physics, EURATOM-CEA/CEN/EUR 5/27 e, EURATOM, Luxembourg, 1974) showed that an indented plasma enhanced plasma stability against localized interchange modes. While the majority of design studies have been performed at low .beta., it has also been known that at very large .beta., there exists a region of operation where stability to ballooning modes could be regained because of the magnetic well effects produced by the large outward Shafranov shift (e.g., Coppi et al., Nucl. Fusion 19, 715 (1979)). This stable region was called the "second region" of stability and many unsuccessful attempts were made to discover operating scenarios which would make this region accessible from the low-.beta. regime. [By accessibility, we mean a demonstration that a method of operation of the device is possible whereby the .beta. (or pressure) of the device can be increased continuously from zero to a very large .beta. value without passing through the unstable region.] For example, detailed numerical calculations (e.g., Monticello et al., Sherwood Meeting, Austin, Tex., April, 1981) demonstrated that in plasmas with nearly circular cross sections the second stable region occurred only for large aspect ratio configurations and accessibility was not possible. In addition, the internal kink has been shown to be a prime candidate responsible for enhanced fast-ion loss through "fishbone oscillations", thus limiting the ability to increase .beta.. It is therefore an object of the present invention to provide a method and apparatus for forming a magnetically confined plasma. Another object of the present invention is to provide a method and apparatus for forming a magnetically confined plasma and avoiding plasma MHD instabilities which defeat plasma confinement. Yet another object of the present invention is to provide a method and apparatus for forming a plasma with an increased beta. Another object of the present invention is to provide a method and apparatus which makes accessible the second region of stability against ballooning modes. Still another object of the present invention is to provide a method and apparatus for forming a high-beta plasma having stabilized ballooning and internal-kink modes thereby minimizing fast ion losses. Additional objects, advantages, and novel features of the invention will be set forth in part in the description which follows, and in part will become apparent to those skilled in the art upon examination of the following or may be learned by practice of the invention. SUMMARY OF THE INVENTION The present invention is the first demonstration for toroidal magnetic confinement devices that the second region of stability against ballooning modes can be accessed with controlled operation. Indeed, under certain modes of operation, it has been found the first and second stability regions may be joined together. Accessing the second stability region is essential to obtaining the high beta necessary for commercial fusion reactors. The present invention also demonstrates the ability to simultaneously achieve complete stabilization to the internal (n=1) kink modes. For toroidal confinement devices, the second region of stability may be accessed by the following scenario: first, modifying the shape of the plasma until it has a bean-shaped poloidal cross-section (by bean-shaped we mean indented or inwardly concave at the inner-major-radius side). Second, operating the device in the first region of stability while further indenting the small-major radius side of the cross-section. When the device is being operated in the first region of stability, .beta. must be kept below the threshold for instability. There are several ways .beta. can be maintained below the threshold for instability, for example: controlling plasma pressure, p, while keeping magnetic field, B, constant; controlling magnetic field B while maintaining plasma pressure constant; or some combination of both. As the indentation is increased, a critical value is reached. This critical value indicates the point at which the second stability region is accessed. Third, after the second stability region has been accessed, .beta. can be increased significantly, to well over 20%. Another feature of the second region of stability is the fact that once large .beta.s are attained, the indentation can be relaxed. An alternate method of accessing the second region of stability is as follows. First the magnetic field would be applied to the device. Then the bean-shaped plasma would be formed. The bean-shaped cross-section would be chosen such that the indentation is at least as large as the critical value. Then the beta would be increased to the desired value for operation. Since operation is in the region where both first and second regions are joined, there are no problems with balloon instabilities. After the desired beta is attained (such as by heating the plasma) then the indentation can be relaxed. The method of the present invention has been demonstrated to provide stability against ballooning modes and against internal kink modes. A theoretical analysis of the stability against ballooning modes is contained in M. S. Chance et al., "Ballooning Mode Stability of Bean-Shaped Cross Sections for High-.beta. Tokamak Plasmas", Phys. Rev. Lett. 51 1963 (November 1983), which is incorporated herein by reference. A theoretical analysis of the stability against internal kinks is contained in J. Manickam et al., "Stability of n=1 Kink Modes in Bean-Shaped Tokamaks", Phys. Rev. Lett. 51, 1959 (November 1983), which is incorporated herein by reference. It has also been found that for the method of this invention, the Mercier modes are even more stable than in conventional tokamaks (especially with the strong minimum-B property of the bean shaping) and the gross stability of the external kink modes is similar to those of conventional tokamaks.
	claims	1. A process for treating radioactive isotopes in liquid, the method comprising:mixing glass beads comprising sodium, calcium and boron with a potassium phosphate solution, wherein the sodium, calcium, and boron react with the potassium phosphate solution to form a hydroxyapatite layer that covers the glass beads;placing hydroxyapatite microspheres comprising the glass beads covered with the hydroxyapatite layer in an ion exchange column;passing the liquid with radioactive isotopes through the ion exchange column of media comprising the hydroxyapatite microspheres for capturing one or more of the radioactive isotopes from the liquid, wherein passing the liquid with radioactive isotopes through the ion exchange column comprises passing said liquid through a dip tube oriented in a generally vertical direction and into a distribution ring attached to the dip tube and suspended within the ion exchange column in a generally horizontal orientation, and wherein the distribution ring comprises a plurality of distribution holes that direct the liquid radioactive waste in a downward direction towards a bottom end of the ion exchange column, whereby as the liquid passes through and near the media contained within the ion exchange column, the one or more radioactive isotopes are separated from the liquid and retained within the ion exchange column by the media; andheating the media and separated radioactive isotopes to form a vitrified waste product, wherein the vitrified waste product contains at least one of the the dip tube and the distribution ring. 2. The process of claim 1, wherein the one or more radioactive isotopes comprise strontium. 3. The process of claim 2, wherein the media include glass-based microspheres. 4. The process of claim 1, wherein the media include Herschelite. 5. The process of claim 1, wherein the media include a surfactant-modified zeolite. 6. The process of claim 1, wherein the liquid with radioactive isotopes is passed through multiple columns, each column containing media for capturing at least one radioactive isotope different from one or more radioactive isotopes separated in the other columns. 7. The process of claim 1, wherein the potassium phosphate solution also comprises potassium hydroxide. 8. A process for treating radioactive isotopes, the process comprising:pyrolizing solid radioactive waste;expelling gas and vapor that result from pyrolizing the solid radioactive waste, wherein the expelled gas and vapor comprise one or more radioactive isotopes;condensing the expelled gas and vapor to form liquid radioactive waste;mixing the liquid radioactive waste with a plurality of microspheres contained in an ion exchange column to transfer one or more radioactive isotopes from the liquid radioactive waste to the plurality of microspheres;responsive to mixing the liquid radioactive waste, pass the waste through a dip tube oriented in a generally vertical direction and into a distribution ring attached to the dip tube and suspended within the ion exchange column in a generally horizontal orientation, and wherein the distribution ring comprises a plurality of distribution holes that direct the liquid radioactive waste in a downward direction towards a bottom end of the ion exchange column;pumping the liquid waste out of the column; andheating the plurality of microspheres with the transferred radioactive isotopes while located inside of the column to form a vitrified waste product, wherein the vitrified waste product contains at least one of the dip tube and the distribution ring. 9. The process of claim 8, wherein the ion exchange column includes a graphite inner layer that acts as a susceptor for inductive heating of the plurality of microspheres. 10. The process of claim 8, wherein the plurality of microspheres comprise glass beads mixed with a potassium phosphate solution to form hydroxyapatite microspheres. 11. The process of claim 8, wherein pyrolizing the solid radioactive waste comprises forming molten radioactive material, and wherein the process further comprises vitrifying the molten radioactive material separately from the plurality of microspheres. 12. The process of claim 11, wherein the ion exchange column is configured to separate and isolate radioactive isotopes that are not vitrified with the molten radioactive material.
054715148	abstract	A fuel assembly for a light-water nuclear reactor includes a plurality of vertical fuel rods (12) which are arranged, in spaced relationship in the lateral direction, between a bottom tie plate (13) and a top tie plate (14). The bottom tie plate and the top tie plate are provided with through-holes (17a) for inlet and outlet of coolant for the fuel rods. Below the bottom tie plate, in the flow path of the water, a debris catcher (18) is arranged. The debris catcher includes at least one helical spring (19) which is fixed in a holder permeable to the coolant and defining at least one slot which is formed as a plane spiral (30) or several concentric annular slots (23, 27) in which slot/slots the helical spring/helical springs is/are arranged.
053176154	claims	1. An exposure apparatus for exposing a workpiece to a pattern of an original with X-rays, said apparatus comprising: a masking device having a movable blade for variably defining an aperture to selectively block and transmit the X-rays to define on the workpiece a desired exposure zone corresponding to said aperture, said movable blade having a window; and detecting means comprising a source of alignment light for detecting a positional relationship between the original and the workpiece by using alignment light having a wavelength longer than the X-rays passing through said window of said movable blade, wherein said source of alignment light is displaced with said movable blade. providing a masking device having a movable blade for variably defining an aperture to selectively block and transmit the X-rays to define on the workpiece a desired exposure zone corresponding to the aperture, wherein said movable blade has a window; positioning the movable blade to set the aperture to be used for the exposure of the workpiece; detecting a positional relationship between the original and the workpiece by using alignment light having a wavelength longer than the X-rays passing through the window of the movable blade, wherein a source of the alignment light is displaced with the movable blade; and exposing a zone of the workpiece corresponding to the set aperture of the masking device with the X-rays passing through the set aperture. a masking device having a movable blade for variably defining an aperture to selectively block and transmit the X-rays to define on the workpiece a desire exposure zone corresponding to the aperture, said movable blade having a window; a light source for providing alignment light having a wavelength longer than the X-rays, for detection of a positional relationship between the original and the workpiece, wherein said source of alignment light is displaced with said movable blade; and a detector for receiving the alignment light provided by said light source and passing through said window of said movable blade, to detect the positional relationship between the original and the workpiece; wherein said movable blade is disposed between the workpiece and said detector with respect to a path of the X-rays. providing a masking device having a movable blade for variably defining an aperture to selectively block and transmit the X-rays to define on the workpiece a desired exposure zone corresponding to the aperture, wherein the movable blade has a window; positioning the movable blade to set the aperture to be used for the exposure of the workpiece; detecting a positional deviation between the original and the workpiece by using alignment light having a wavelength longer than the X-rays, passing through the window of the movable blade, wherein a source of the alignment light is displaced with the movable blade; and exposing a zone of the workpiece corresponding to the set aperture of the masking device, with the X-rays passed through the set aperture. 2. An apparatus according to claim 1, wherein said movable blade blocks the X-rays. 3. An apparatus according to claim 1, wherein the alignment light comprises a laser beam. 4. An apparatus according to claim 1, wherein said window is made of a material effective to block the X-rays and to transmit the light, and wherein said detecting means projects the alignment light through said window to detect the positional relationship. 5. An apparatus according to claim 1, wherein said movable blade has as portion made of a metal, and wherein said apparatus further comprises temperature control means for controlling temperature of said metal portion. 6. An apparatus according to claim 1, further comprising position control means for controlling a position of said movable blade on the basis of detection by said detecting means. 7. An apparatus according to claim 1, further comprising moving means for moving the workpiece relative to the original, wherein said moving means is operable to move the workpiece so as to sequentially bring different portions of the workpiece to a position below the original to effect step-and-repeat exposures. 8. An apparatus according to claim 1, wherein said masking device is four movable blades corresponding to four sides of the exposure zone, each being movable rectilinearly and reciprocatingly. 9. An apparatus according to claim 1, wherein said detecting means optically detects the positional relationship between an alignment mark provided on the workpiece and an alignment mark provided outside of the exposure zone on the original. 10. An apparatus according to claim 1, wherein said movable blade is made of a material effective to block the X-rays but to transmit the second alignment light. 11. An apparatus according to claim 1, wherein said movable blade is disposed between said detecting means and the workpiece, with respect to a path of the X-rays. 12. An exposure method for exposing a workpiece to a pattern of an original with X-rays, said method comprising the steps of: 13. A method according to claim 12, wherein the alignment light comprises a laser beam. 14. A method according to claim 12, wherein the positional deviation is detected by projecting the alignment light through a portion of the blocking member which portion is effective to block the X-rays but to transmit light. 15. A method according to claim 11, wherein the movable blade is disposed between the workpiece and detecting means provided for said detection, with respect to a path of the X-rays. 16. An exposure apparatus for exposing a workpiece to a pattern of an original with X-rays, said apparatus comprising: 17. An apparatus according to claim 16, wherein the alignment light comprises a laser beam. 18. An apparatus according to claim 16, wherein said movable blade is made of a material effective to block the X-rays but to transmit the second light. 19. An apparatus according to claim 15, wherein said movable blades have a portion made of a material effective to block the X-rays but to transmit the light, wherein said light source projects the alignment light through the window while the detector receives the light passing through said window. 20. An apparatus according to claim 16, wherein said movable blades block the X-rays. 21. A method of manufacturing semiconductor devices by exposing a workpiece to a pattern of an original with X-rays, said method comprising the steps of: 22. A method according to claim 21, wherein said movable blade providing step comprises the step of disposing the masking device between the workpiece and a detecting means provided for said detection, with respect to a path of the X-rays.
059498366	description	DETAILED DESCRIPTION OF THE INVENTION The features and other details of the apparatus and method of the invention will now be more particularly described with reference to the accompanying drawings and pointed out in the claims. The same number present in different figures represents the same item. It will be understood that the particular embodiments of the invention are shown by way of illustration and not as limitations of the invention. The principle features of this invention can be employed in various embodiments without departing from the scope of the present invention. The specific activity of a radioisotope, within a volume of target material, is the number of radioactive disintegrations per second of nuclides of the radioisotope (in curies (Ci)) measured per gram of the radioisotope's element, including all isotopes of the element, within the volume of target material. Specific activity provides an indication of the concentration of the radioisotope within the volume of target material. Typically, the specific activity is not uniform across a volume of target material, but is averaged across the volume of target material. The level of specific activity, which constitutes a high specific activity, is dependent upon the radioisotope and its use. For example, wherein the radioisotope is molybdenum-99 (Mo.sup.99), which subsequently decays to the daughter product technetium-99 (Tc.sup.99), a high specific activity for Mo.sup.99 is typically an average specific activity of about 0.5 Ci/gram of molybdenum, or more. Preferably, the specific activity of Mo.sup.99 is about 1.0 Ci/gm, or more. More preferably, a high specific activity of Mo.sup.99 is about 5 Ci/gram, or more. Even more preferably, a high specific activity of Mo.sup.99 is about 10 Ci/gram, or more. A radioisotope can be generated in a target material using high energy photons from a photon beam in at least one isotopic conversion reaction. A target material is a material which consists of or contains a targeted isotope, which when exposed to high energy photons, forms the radioisotope as a product. Typically, a targeted isotope has a high atomic number (Z), for example, a Z of about 30 or more. A radioisotope product can be a final product, such as Cadmium-115 or Tantalum-179. Alternatively, a radioisotope product, such as Cadmium-109 or Osmium-191, can be an intermediate which subsequently decays to form a desired daughter product. Preferably, a radioisotope product is longer-lived. A longer-lived radioisotope, as defined herein, is a radioisotope with a half-life suitable to allow shipping and the subsequent use of the radioisotope, or a daughter product, after generating the radioisotope. Typically, a longer-lived isotope has a half-life of about 12 hours or more. Preferably, the half-life is about 48 hours or more. More preferably, the half-life is about 60 hours or more. Most preferably, the radioisotope product is Mo.sup.99. Suitable isotopic conversion reactions include, for example, (.gamma.,n), (.gamma.,2n), (.gamma.,p) and (.gamma.,pn) reactions. An energy level, suitable for a high energy photon, is an energy level which is at least equal to the threshold (minimum) energy level, of the Giant Resonance region of the cross-section versus energy curve for the desired isotopic conversion reaction, required to produce the reaction between a photon and the targeted isotope. The specific activity of a photon-beam generated radioisotope, within a volume of a target material, depends upon several variables, including the intensity (photon energy per unit area per unit time) of the high energy photons in the photon beam and the thickness of the target material. As shown in FIG. 1, the peak specific activity level, for a photon beam of any intensity, is at the target material surface irradiated by the photon beam. A photon beam with a higher intensity of high energy photons, irradiating the same target material, typically generates a higher peak specific activity than does a photon beam with a lower intensity of high energy photons. A high intensity of high energy photons is an intensity sufficient to generate a high specific activity of a radioisotope. Typically, a suitable intensity of high energy photons is at least 50 microamps/cm.sup.2 (.mu.a/cm.sup.2). Preferably, the intensity of high energy photons is at least 500 .mu.a/cm.sup.2. More preferably, the intensity of high energy photons is at least 1,000 .mu.a/cm.sup.2. In addition, as also shown in FIG. 1, specific activity levels within the target material decrease exponentially with increasing depth along the thickness of the target material. The thickness of the target material is the distance from the irradiated side of the target material to the opposite face. Thus, the average specific activity of a radioisotope within a volume of target material increases with decreasing target material thickness. The maximum specific activity (saturation activity) achievable by isotopic conversion in a volume of target material varies linearly with the production rate of the radioisotope. Typically, saturation activity is achieved only following irradiation periods that are significantly longer than the half-life of the radioisotope. Saturation activity (S) is calculated by the following equation: EQU S=1.62.times.10.sup.13 f.multidot.R/A wherein f is the fraction of isotope of the targeted element which is targeted isotope and A is the atomic weight of the targeted element. R, which is indicative of the intensity of high energy photons, is the photon path length per unit volume and per unit energy (".phi.(E)") weighted by the photon cross-section (".sigma.(E)"), integrated over all photon energy levels. The specific formula for calculating the value of R is as follows: EQU R=.intg..sigma.(E).multidot..phi.(E).multidot.dE. The photon energy levels included in the calculation of R may be limited to those in the Giant Resonance range as lower energy photons are not effective. Specifically, lower energy photons do not result in photonuclear conversion of Mo.sup.100 to Mo.sup.99. One embodiment, of the apparatus for producing a high specific activity of a product radioisotope in a volume of target material, is illustrated in FIG. 2. Apparatus 10 includes target material 12, convertor 14 and electron accelerator 16. Target material 12 contains a loading of a targeted isotope which can be established based upon the intended isotopic conversion reaction and the concentration of product radioisotope desired. The specific isotopic conversion reactions occurring within target material 12 typically depend upon the desired product isotope and the availability of nuclei of the targeted isotope within target material 12. In one embodiment, the loading of a targeted isotope in target material 12 is at naturally occurring levels. Preferably, target material 12 contains enriched levels of the targeted isotope. The targeted isotope can be in elemental form, in at least one compound (e.g., a salt or oxide), and/or complexed. The targeted isotope within the target material can be in any physical state, for example, a particulate, a liquid, in solution, in a suspension, in a slurry, or a in a larger solid mass. Examples of other components optionally contained in target material 12 include materials in which the targeted isotope is retained, such as a metallic or ceramic material, or materials in which the targeted isotope is dispersed such as in a liquid (e.g., water or oils) or in particulates. Apparatus 10 further includes electron beam 18 and photon beam 20. Electron beam 18 is generated by electron accelerator 16 and is directed into convertor 14, wherein photon beam 20, which includes high energy photons, is generated. Photon beam 20 radiates from convertor 14 into target material 12. Typically, photon beam 20 is a substantially collimated high energy photon beam. A suitable convertor contains at least one high Z material, for example tungsten or platinum, which is refractory under the conditions of the method of invention. A high Z material is used to improve the efficiency of the conversion within convertor 14 of high energy electrons from electron beam 18 into high energy photons to form photon beam 20. The total extent of convertor 14 in the direction of the trajectory of electron beam 18 should be sufficient to absorb a significant portion of the energy of electron beam 18 while transmitting photon radiation in an energy range suitable for the desired isotopic conversion reaction. Concurrent with transforming the energy of electron beam 18 into high energy photons in photon beam 20, convertor 14 also shields target material 12 from any significant residual electron beam. If convertor 14 is too thick, photons emitted from convertor 14 will be degraded in energy due to passing through the material of convertor 14. If convertor 14 is too thin, significant levels of electrons will pass through convertor 14 and impinge upon target material 12. The preferred thickness of convertor 14, for obtaining optimum product isotope yield, depends on electron beam energy, the composition of convertor 14, and the Giant-Resonance region threshold energy of the targeted isotope. An example of an optimal convertor is a convertor containing approximately six plates of tungsten alloy of aggregate thickness 5 mm separated by cooling ducts for water cooling. The intensity of high energy photons generated in convertor 14 is proportional to the power density (PD) of electron beam 18 in convertor 14. Thus, the specific activity of a radioisotope within a volume of target material 12 is also proportional to the power density. Power density within convertor 14 is calculatable by the following equation: EQU PD=E.times.i/V wherein E is the energy of electron beam 14, i is the current of electron beam 18 and V is the volume of convertor 14 through which electron beam 18 passes. The power density used in this invention is limited by the heat removal capacity of convertor 14. In another embodiment illustrated in FIG. 3, convertor 14 is composed of two or more plates 22 of high Z material, such as tungsten, instead of a single solid convertor to allow better heat removal from convertor 14 and thus, higher power densities of electron beam 18 therein. Plates 22 can be fabricated from the same or different material. The plates are typically enclosed by external shell 24, which maintains the geometry of convertor 14 and also retains any optional coolant within convertor 14. In a preferred embodiment, plates 22 do not have equal thicknesses. The thicknesses of the plates is varied to equalize the heat loads on the plates. The heat load on each plate is derived from the energy transferred to the plate by electron beam 18 and by generated photons passing through each plate. Typically, the heat loads on plates distal to electron accelerator 16 are greater than the heat loads on proximal plates as electron beam 18 deposits energy in a plate after the electrons are slowed by previous plates. In addition, photons generated in the proximal plates can also deposit energy in subsequent, distal plates. Thus, in a more preferred embodiment, plates 22 proximal to electron accelerator 16 are thicker than plates 22 which are distal to the electron accelerator 16 to better equalize the heat generation in each plate 22. Plates 22 and cooling channels 26 in convertor 14 do not need to be perpendicular to the direction of electron beam 18. Preferably, the cross-sectional areas of convertor 14, or plates 22, are perpendicular to the path of electron beam 18. Optionally, means are provided for removing heat from at least a portion of convertor 14. Heat removal is provided by typical means, such as by radiation, conduction and/or convection. Heat removal means are disposed around and/or through convertor 14. Examples of suitable heat removal means include coolant channels 26 which are disposed within the material forming convertor 14 (e.g., wherein the convertor material is a honeycomb), etched along the surface of convertor 14, etched along the surface of plates 22 and/or are disposed between plates 22. Alternatively, convertor 14 includes porous material in the form of frit wherein coolant flows through the interstices within the frit for heat removal. Heat removal means also include convertor inlet 28 and convertor outlet 30, which are disposed at shell 24 of convertor 14. Preferably, heat generated within convertor 14, or within each plate 22 of convertor 14, is removed by fluid coolant flow into convertor 14 through convertor inlet 28, through coolant channels 26 and out of convertor 14 through convertor outlet 30. Suitable means of fluid coolant flow include, for example, single-pass fluid flow, natural circulation and forced recirculation. Typically, outside of convertor 14, the coolant is then cooled, such as by being directed through heat exchanger 32A. Suitable fluid coolants include liquids, such as water or liquid gallium and gases, such as helium. For very high power densities within convertor 14, such as greater than about 3 thousand watts/cm.sup.3 or more, it is preferred that convertor 14 be a porous metallic frit which is cooled by fluid coolant flowing at high pressure through the pores, or interstices, within the frit. In the embodiment wherein convertor 14 is tungsten and the targeted isotope is Mo.sup.100, the optimum yield of a Mo.sup.99 product isotope yield is when plates 22 of convertor 14 have a combined thickness slightly less than the stopping distance for an electron in electron beam 18. When plates 22 have a combined thickness less than the electron stopping distance, backing 34 is disposed between convertor 14 and target material 12 to capture electrons without significantly degrading the energy of the photon beam. Suitable materials for backing 34 include lower Z metals such as aluminum. Typically, the high energy photon beam is directed through backing 34 at or near the center of backing 34. Further, the cross-sectional area of backing 34 is preferably equal to or larger than the width of high energy photon beam 18. Optionally, backing 34 can be cooled by means for removing heat, not shown, such as heat transfer to a cooling medium (e.g., water). In yet another embodiment illustrated in FIG. 4, convertor 14 consists of molten or liquified high Z material 33, which is recirculated from convertor inlet 28, through convertor 14, out of convertor outlet 30, through heat exchanger 32B, and subsequently back into convertor inlet 28. Heat generated in convertor material 33 within convertor 14 by the electron beam then dissipates, or is removed by suitable means, such as heat exchanger 32B, while the convertor material is outside of the convertor. FIG. 5 illustrates an alternative embodiment of the apparatus of this invention wherein separate, or separable, increments of target material 12 are irradiated in series thereby producing a high specific activity of radioisotope in the first increment and pre-irradiating the second increment to commence building up the concentration of the radioisotope within the increment. Apparatus 100 includes target assembly 36, convertor 14 and electron accelerator 16. Electron beam 18 is generated by electron accelerator 16 and is directed into convertor 14, wherein photon beam 20, which includes high energy photons, is generated. Photon beam 20 extends from convertor 14 into target assembly 36. Target assembly 36 includes a target material which is separated or separable into at least two increments, with first target material increment 38 located proximal to convertor 14 and second target material increment 40 located adjacent to first target material increment 38 and distal to converter 14. Additional targets material increments 42 are disposed, in series, behind second target material increment 40. An increment of a target material is an amount of target material which is separate or separable from the target material contained within target assembly 12. Each increment of target material, such as first target material increment 38, second target material increment 40 and additional target material increments 42, contains a loading of a targeted isotope within the target material of the target. Typically, wherein the targeted isotope is contained within a larger solid mass, first target material increment 38 and second target material increment 40 consist of separate sections of the target material. Target assembly 36 also includes inlet 44A and outlet 46A. Inlet 44A is disposed at or near the end of target assembly 36 distal to convertor 14. Inlet 44A is provided as a means for directing additional targets 21 into target assembly 36 on the distal side of second target material increment 40. Outlet 46A is disposed at or near the end of target assembly 36 that is proximal to convertor 14. Outlet 46A is provided as a means for separating a distal target material increment from its adjacent target material increment (e.g., separating first target material increment 38 from second target material increment 40) by directing the distal target material increment out of target assembly 36 through outlet 46A. Preferably, target assembly 36 also includes means, such as pushrod 48, for conveying increments of target material through target assembly 36 toward convertor 14, and then out of target assembly 36. Alternatively, other known means for non-destructively conveying target material can also be used to convey targets or target material through target assembly 36. Examples of other suitable conveying means include, for instance, conveyor belts, screws, pistons and pumps. The target assembly 36 may further include photon reflector 50. Photon reflector 50 is disposed around at least a portion of target assembly 36. Photon reflector 50 is typically composed of high Z metals (e.g., a Z of about 30 or more), such as molybdenum-98, uranium, tantalum, tungsten, lead and other heavy metals. Photon reflector 50 reflects at least a portion of the high energy photons impinging the reflector material (e.g., from the incoming photon beam or scattered from the in-series target material increments) into the target material within target assembly 36. Optionally, target assembly 36 includes neutron shielding 52 which is disposed at least partially around photon reflector 38. Suitable types of neutron shielding include shielding with a high hydrogen content, such as a plastic or water, which thermalizes and/or captures at least a portion of the neutrons emitted during an isotopic conversion reaction. The depth of target material 12 through which photon beam 20 passes within the aggregate of in-series target material increments, disposed within target assembly 36 is determined based upon the loading of targeted isotopes within each increment, the desired concentration of product isotopes within each increment, the energy level of photon beam 20 and the period of irradiation. Preferably, the target material, contained in the in-series target material increments, has an aggregate thickness that results in the capture of all but an insignificant amount of high energy photons in photon beam 20 which impinge the target material and do not scatter outside of the target material. For example, wherein the targeted isotope is Mo.sup.100 and the desired product is Mo.sup.99, the aggregate thickness of the targets is typically between about 6 cm to about 10 cm for a photon beam produced by a tungsten convertor exposed to a 30-40 Mev electron beam. The cross-sectional area of target material 12 within target assembly 36 perpendicular to photon beam 20 can be varied depending upon the focal area of photon beam 20 on target material increment 38 and the expected spread of the photon beam 20 along the path of photon beam 20 through target material 12. The cross-sectional area of target material 12 is usually about equal to, or larger than, the focal area of photon beam 20. In an alternative embodiment illustrated in FIG. 6, target material 12 is in a particulate, liquid, slurry or any other physical form wherein an increment of target material 12 is not contained in a single solid mass. Thus, increments of target material 12 are not separate but are separable. Target assembly 36 includes means for containing target material 12 within target assembly 36, such as cylinder 54 which is disposed within target assembly 36. Suitable containing means, include containers for solids and/or liquids, which are refractory, such as titanium. The material composition and structural design of the container should not result in a significant reduction in the energy of photon beam 20 or a significant increase in the scatter of photons from photon beam 20. Cylinder 54 includes baffles 55 which control the flow in cylinder 54 to assure generally uniform irradiation. Target assembly 36 also includes means for directing increments of target material 12 through cylinder 54. This directing means includes inlet 44B and outlet 46B. Inlet 44B is disposed at or near the end of cylinder 54 distal to convertor 14. Outlet 46B is disposed at or near the end of cylinder 54 that is proximal to convertor 14. In this embodiment, target material 12, which is typically in liquid, slurry or particulate form, is directed into cylinder 54 through inlet 44B, moves towards and the proximal end of cylinder 54, and then comes out of cylinder 54 through outlet 46B. The movement (e.g., flow) of target material 12 through cylinder 54 can be continuous of intermittent. Suitable means to direct flow of target material 12 include, for example, pumps, pistons and gravity feeding. The flow of target material 12 through cylinder 54 can be controlled, for instance, by a valve or clamp located in a position suitable to stop flow (e.g., at inlet 44B or outlet 46B) and/or by controlling the flow directing means (e.g., starting and stopping a pump). In another embodiment illustrated in FIG. 7, wherein the increments of target material 12 are separate, but not solid masses, target assembly 36 further includes means for separately containing each increment of target material 12. Typically, target material 12 is in a particulate, liquid or slurry form. Suitable containing means, such as container 56, include containers which can contain a solid and/or liquid, wherein the container is refractory under the method of this invention. The material composition and structural design of the container should not result in a significant reduction in the energy of photon beam 20 or a significant increase in the scatter of photons from photon beam 20. An example of a suitable container material is titanium. In this embodiment, containers 56 enter the distal end of target assembly 36 through inlet 44B, are directed toward the proximal end of target assembly 36 while concurrently being irradiated by photon beam 20, and then leave target assembly 36 through outlet 46B. Operation of the embodiment of FIG. 2 for producing a high specific activity of a radioisotope will now be described. Electron accelerator 16 generates electron beam 18 which is directed into convertor 14. At least a portion of the electrons of electron beam 18 are captured in an (electron, .gamma.) reaction by the high Z material of convertor 14 to generate photons, including high energy photons in photon beam 20. Typically, most electrons are captured and most photons pass through convertor 14. Typically, electron accelerator 16 generates an electron beam 18 with an average energy level of about 25 MeV or more, preferably between about 30 MeV and about 50 MeV. The total power of electron beam 18 is limited by the design of electron accelerator 16 and by the design, thickness and heat removal capability of convertor 14. If the beam energy is too low, there will not be sufficient photons in the Giant Resonance region to produce a high specific activity of the radioisotope and the electron range in convertor 14 will be so short as to make heat removal from convertor 14 very difficult. If the beam energy is too high, many photons will have energies above the optimal range, direct electron heating of target material 12 will be a problem and electron accelerator 16 will be relatively expensive. In addition, increased production of impurities, such as niobium, can result for other isotopic conversion reactions. Photon beam 20 is directed from convertor 14 and focused onto target material 12. Target material 12 is typically placed in close proximity to convertor 14 and in alignment with the exit of photon beam 20 from convertor 14. Sufficient distance between convertor 14 and target material 12 may be left to interpose material to attenuate electromagnetic fields to deflect electron beam 18 or to interpose material to modify the photon spectrum of photon beam 20, but this distance is minimized in order to use the photon beam at high intensity. If no attenuation is required, target material 12 may be in contact with convertor 14. Within target material 12, at least a portion of the high energy photons of photon beam 20, react with the targeted isotope to form a concentration of a radioisotope within the target material by an isotopic conversion reaction, such as by (.gamma.,n), (.gamma.,2n), (.gamma.,p) or (.gamma.,pn) reaction. Preferably, a significant number of the photons of photon beam 20 are high energy photons which have an energy level falling within the range of energy levels included in the Giant Resonance region of the cross-section versus energy curve for the desired isotopic conversion reaction. More preferably, a significant portion of the photons of photon beam 20 have energy levels about equal to the peak energy level of the Giant Resonance region. For heavier materials, the energy levels corresponding to the Giant Resonance region are relatively lower while for lighter materials the energy levels are relatively higher. Preferably, the energy of electron beam 18 should be about 2 to about 3 times the energy level of the peak of the Giant Resonance region of the targeted isotope. For example, in the (.gamma.,n) isotopic conversion of Mo.sup.100 to Mo.sup.99 it is preferred that at least a significant portion of photons in photon beam 20 have energy levels falling within the Giant Resonance region for this reaction, specifically between the threshold energy level of about 10 MeV and the high energy limit of about 19 MeV. More preferably, photon energy levels are about 15 MeV, which is the peak of the Giant Resonance region. The electron beam energy for this isotopic conversion is typically between about 25 Mev to about 50 Mev, with a preferred energy range of about 35 Mev to about 40 MeV. The energy level of a generated photon is directly dependent upon the energy level of electron beam 18, with the peak energy level of generated photons being equal to about the energy level of electron beam 18. Typically, most generated photons have energy levels at less than half the peak energy. Therefore, the energy level of at least a portion of the electrons in electron beam 18 at a minimum must be equal to the threshold (minimum) energy level required to produce the desired isotopic conversion reaction between a generated photon and the targeted isotope. Preferably, the energy level of electron beam 18 is within or above the Giant Resonance region of the desired isotopic conversion reaction. In a preferred embodiment, wherein the targeted isotope is molybdenum-100 (Mo.sup.100) which is isotopically converted to molybdenum-99 (Mo.sup.99), which then decays to the desired daughter product technetium-99 (Tc.sup.99), the photon beam produced includes .gamma. radiation at an energy level of about 8 Mev or more. More preferably, a substantial amount of the .gamma. radiation produced is at energy levels between about 8 Mev and about 16 MeV. Achievement of an average specific activity of Mo.sup.99 of about 1.0 Ci/gm of Mo in solid molybdenum requires a relatively high power density in convertor 14. Specifically, in the saturation activity equation, the product of f.multidot.R must have a value greater than about 2.2.times.10.sup.-8 sec.sup.-1. This value of R is difficult to achieve because of technical limitations on electron beam power density and convertor heat removal. Therefore, the volume in which the average specific activity of 1.0 Ci/gm can be maintained is typically limited to target material volumes having relatively small thicknesses. In determining the maximum volume of target material, the cross-sectional area of the target material usually must be equal to or less than the focal area of photon beam 20. Thus, target material volume is often limited to a few cubic centimeters or less. For example, for a natural molybdenum target, containing approximately 10% Mo.sup.100, a 35 Mev electron beam of 1.0 milliampere current focused onto a 1.0 cm radius target disk yields, with an optimal convertor, an average specific activity of about 1.0 Ci/gm for a target material thickness of about 0.5 cm. The power density in the active regions of the convertor would be about 35,000 watts/cm.sup.3. Higher specific activities can be achieved by isotopic enrichment of the target material. A target material enriched to 100% Mo.sup.100 would yield a specific activity in excess of 10 Ci/gm up to a target material thickness of about 0.5 cm for the same conditions. Molybdenum target thicknesses greater than 0.5 cm, having an average specific activity of at least 1.0 Ci/gm, can be obtained by varying the isotopic enrichment of Mo.sup.100 in the target material and/or by varying the energy levels of the photons in the photon beam, providing the value of the product f.multidot.R is at least 2.2.times.10.sup.-8 sec.sup.1. For a thick target, the activity produced in the first 0.5 cm depth of the target is only 28% of the total generated in the target. However, the other 72% of the desired product isotope is so diluted with unconverted target material as to be below commercial interest. On the other hand, to irradiate a single target of 0.5 cm thickness or less results in lost photon energy. The portion of the thick target with less than threshold activity represents a potentially valuable resource, unusable if unimproved. Accordingly, by providing an incremental target as in FIG. 5, only that portion of the target which has been irradiated to an average specific activity above a given threshold value is removed for processing. Additional portions of the target, irradiated to less than the threshold value, can be sequentially irradiated to the threshold value in such fashion as to optimize the combination of specific activity of individual target elements and total radioisotope production rate. Preferably, each target increment is 0.5 cm thick or less. Within, at least, first target material increment 38 and second target material increment 40, a portion of the high energy photons of photon beam 20, react with the targeted isotope to form a high specific activity in first target material increment 38 and pre-irradiate second target material increment 40, and possibly additional target material increments 42, to commence building up the specific activity of the radioisotope within these increments. This method also includes moving first target material increment 38 and second target material increment 40 toward outlet 46A, and closer to convertor 14, by the action of push rod 48 applying force to the distal side of second target material increment 40 through additional target material increments 42. Alternately, the targets can be moved by any suitable automated or non-automated means. Further, the movement of targets can be continuous, concurrent, sequential or stepwise. Ultimately, first target material increment 38 is pushed through outlet 46A and is removed from target assembly 36. Further second target material increment 40 is pushed to the original position of first target material increment 38 whereupon photon beam 20 then focuses upon second target material increment 40 to complete producing a high specific activity therein. Additional target material increments 42 can be added in-series behind second target material increment 40 through inlet 44A. In this method, the ratio of the specific activity of the product radioisotope in each increment to the amount of product isotope removed per unit time can be optimized depending upon the need for a high discharge rate of product radioisotope or a high specific activity of product radioisotope. The concentration of the product radioisotope generated by the isotopic conversion reaction is dependent upon the intensity of the high energy photons in photon beam 20, upon the volume of target material 12 irradiated, upon the radioactive half-life of the product isotope, and upon the amount of target material 12 which is irradiated. The intensity of photons is approximately dependent linearly upon the current level of electron beam 18 for the same focal area, with higher currents generating more high energy photons per unit time, which then are directed into the target material to react with more targeted isotope per unit time. The volume of target material 12 irradiated by photon beam 20 depends upon the focal area of photon beam 20 upon target material 12 and the amount of photon scatter within the target material. Typically, the focal area of photon beam 20 is a function of the angle of emission of high energy photons from convertor 14. Most higher energy photons, having an energy level which falls within the Giant Resonance region for the desired isotopic conversion reaction, are emitted in a narrow cone whose axis is aligned along the direction of an extended axis of electron beam 18. The intensity of higher energy photons, which are emitted at an angle to the axis of the cone, rapidly decreases as the angle from the cone increases. For instance, at an angle of about 5 degrees from the axis of the cone, the intensity of peak photons is about one fifth of the intensity of peak photons emitted about the center of the cone. In addition, the intensity of higher energy photons, having approximately one-half peak photon energy, is lower by about two orders of magnitude at an angle of 25 degrees from the axis of the cone than the intensity along the axis of the cone. Thus, photon beam 20 is strongly peaked in the forward direction along an extended axis of electron beam 18. Therefore, the focal area of photon beam 20 is determined by the focal area of electron beam 18 on convertor 14. With increasing electron beam energies, the focal area of photon beam 20 becomes smaller with a minimum area being the size of the focal area of electron beam 18 on convertor 14. Thus, with increasing photon beam energies, the cross-sectional area of target material 12 is further limited. To optimize the specific activity of product radioisotope in each target material increment, when removed from the target assembly, the focal width of photon beam 20 is minimized to produce a higher concentration of product radioisotope near the center of first target material increment 38 with lower concentrations near the edges of the target. As photon beam 20 travels through the target material and spreads, such as from scattering, the concentration is reduced near the center of the target material and is increased nearer to the edges of the target material 12. Thus, after passing through first target material increment 38, photon beam 20 will pre-irradiate second target material increment 40 and additional target material increments 42 to produce lower levels of product isotope throughout these incremental targets (e.g., near the centers and at the edges). Preferably, the focal area of electron beam 18 is minimized to attain greater concentrations of product isotope near the centers of the targets. The lower limit on focal area of electron beam 18 on convertor 14 is dependent upon the heat dissipation capability of convertor 14. The focal area of electron beam 18 should not be so small as to create a high power density in the affected potion of convertor 14 which leads to localized melting, destruction and/or loss of function of the convertor material. The amount of time a target is irradiated can depend upon the movement rate of the in-series target material increments, while in photon beam 20, toward outlet 46. Target material increments are introduced, moved and discharged at rate such that the combination of segment thickness and discharge rate yields a product of the desired specific activity of product isotope. A high discharge rate of targets will result in the recovery of a larger fraction of the generated radioisotope but the specific activity of the discharged material will not be as high as that which would result, all other factors remaining unchanged, from a low target material increment discharge rate. FIG. 8 further illustrates the calculated effect on production rate and specific-activity of product of varying the flow rate of target material within the photon beam. FIG. 8 is based upon an electron beam energy of 35 MeV, an electron beam current of 1.10 ma, and cylindrical Mo.sup.100 target segments which are 2.0 cm in radius and 0.5 cm thick. The method of this invention can also be employed to produce concentrations of stable isotopes. The invention will now be further and specifically described by the following examples. EXAMPLE 1 Mo.sup.99 Production by Photonuclear Transmutation of Mo.sup.100 A cylinder of molybdenum (4 inches diameter), having a natural isotopic abundance, was sliced in planes perpendicular to the length of the cylinder into separate foils and slabs of molybdenum. Each foil was followed by a separate slab. Each foil had a thickness of about 0.01 inch (0.25 mm), while each slab had a thickness between about 0.75 inches and about 1.5 inches. The foils were used to determine the specific activity of Mo.sup.99 at different points within the aggregate thickness of the foils and slabs. In the target, the six foil/slab units were situated in series, with the slabs closer to the .gamma. beam source having the narrower widths. Each foil or slab was touching the adjacent slab or foil. A 2 inch diameter, 4.3 mm thick tungsten slab, used as a convertor plate, was located between the .gamma. beam source and the target. The convertor was also touching the first foil of the target. A 28 MeV electron beam, having a current of 1.84 microamperes (.mu.a) and a beam width of 1.5 cm, was directed substantially perpendicularly into the side of the convertor proximal to the electron beam source. A .gamma. beam was generated, substantially perpendicular to the distal side of the convertor. The .gamma. beam was directed into the target. The target was exposed for 4.6 hours to the generated .gamma. beam generated. Twenty-six hours after irradiation, the total activity of technetium-99 (Tc.sup.99), and the Giant-Resonance beam half-width, were then measured for each foil using a calibrated intrinsic-germanium crystal, by measuring the amount of .gamma.s having an energy specific to Tc.sup.99 decay (i.e., 140.1 keV) which were emitted at the center point of each foil, and by measuring the radial distance from the center of the foil over which the activity is reduced by one half to show beam spread. The results of center point activity measurements for the six sequential foils are provided in FIG. 9. As shown therein, the activity of Tc.sup.99 measured at the center point of the first foil, located at surface of the target (depth=0), was 30.3 microcurie (.mu.Ci). The center point activities for foils deeper in the target declined non-linearly as a function of their relative depths within the target. This demonstrates that the intensity of the photon flux in the Giant-Resonance energy range falls off quickly with distance in the target material. The half-width measurements for the six sequential foils are also provided in FIG. 9. The half-width of the first foil (depth=0) was 1.5 cm. The half-widths measured for foils deeper in the target showed some increase with depth, for example the half-width for a foil at a depth of about 6 cm was about 3.3 cm. These half-width measurements demonstrate that the .gamma. radiation beam, though spreading from scatter of .gamma.s within the target, remained sufficiently collimated to support the production of Mo.sup.99 throughout a cross-section of the target without a significant loss of .gamma. radiation energy from the target material. EQUIVALENTS While this invention has been particularly shown and described with references to preferred embodiments thereof, it will be understood by those skilled in the art that various changes in form and details may be made therein without departing from the spirit and scope of the invention as defined by the appended claims. Those skilled in the art will recognize or be able to ascertain using no more than routine experimentation, many equivalents to the specific embodiments of the invention described specifically herein. Such equivalents are intended to be encompassed in the scope of the claims.
047215976	claims	1. In a nuclear reactor system requiring periodic physical manipulation of spent fuel rods, the method of compacting fuel rods from a fuel rod assembly comprising the steps of: (1) removing the top end from said fuel rod assembly; (2) passing multiple fuel rod pulling members having electrodes of weld elements in leading ends thereof in sequence through a fuel rod container and thence through respective consolidating passages in a fuel-rod directing chamber; (3) welding said weld elements of said pulling members to the top end of respective said fuel rods corresponding to the respective pulling members; (4) drawing each of said pulling members axially to draw the respective engaged fuel rods in one axial direction through the respective said passages in said chamber to thereby consolidate said fuel rods into a compacted configuration of a cross-sectional area smaller than the cross-sectional area occupied thereby within said fuel rod assembly; and (5) drawing all of said engaged fuel rods concurrently and substantially parallel to one another to said one axial direction into said fuel rod container while maintaining said compacting configuration in a fuel rod density which is greater than that of the fuel rod density of said fuel rod assembly. a weld cap carried on an extended end of the pulling element for welding to the fuel rod, an electrode carried by the pulling element for welding the cap to the fuel rod, said electrode is supported by said pulling element to form an air gap with said cap, and means for delivering an electrical current to said electrode to generate sufficient heat to weld the cap to the fuel rod. 2. The method of claim 1 wherein the fuel rod density in the said fuel rod container is at least twice that of the fuel rod density in the said fuel rod assembly. 3. The method of claim 1 wherein each of the said pulling elements is connected to a common tensioning member. 4. The method of claim 1 wherein the said pulling elements are welded to the said fuel rods by inert arc welding of a cap member connected to the ends of the pulling elements. 5. The method of claim 1 wherein the said fuel rods are merged toward one another as they pass through the said fuel rod directing chamber. 6. A nuclear fuel rod pulling apparatus for pulling a nuclear fuel rod from a fuel rod assembly into a storage container, said apparatus including the combination of a pulling element extending along the direction of travel for the fuel rod, 7. The apparatus according to claim 6 further including means for delivering an inert gas to said weld cap for maintaining an electric arc established between said electrode and the cap submerged in the inert gas. 8. The apparatus according to claim 7 wherein said means for delivering inert gas includes a header connected to an end portion of said pulling element which is remote from said weld cap. 9. The apparatus according to claim 6 wherein said pulling element includes a tube, and a holder carried by said tube for supporting said weld cap. 10. The apparatus according to claim 6 wherein said means for delivering an electrical current includes a holder for supporting said electrode and an electrical conductor connected by said holder to said electrode while extending along said pulling element. 11. The apparatus according to claim 10 wherein said pulling element includes a tube.
	description	This application is based upon and claims the benefit of priority from prior Japanese Patent Application No. 2007-168518 filed on Jun. 27, 2007 in Japan, the entire contents of which are incorporated herein by reference. A lithography technique which advances micropatterning of a semiconductor device is a very important process which generates a unique pattern in semiconductor manufacturing processes. In recent years, with a high integration density of an LSI, a circuit line width required for a semiconductor device are micropatterned every year. In order to form a desired circuit pattern to the semiconductor device, a highly accurate original pattern (also called a reticle or a mask) is necessary. In this case, an electron beam writing technique essentially has excellent resolving power and is used in production of a highly accurate original pattern. FIG. 5 is a conceptual diagram for explaining an operation of a variable-shaped electron beam writing apparatus. The variable-shaped electron beam (EB) writing apparatus operates as follows. In a first aperture plate 410, an oblong (for example, rectangular) opening 411 to shape an electron beam 330 is formed. In a second aperture plate 420, a variable shaping opening 421 to shape the electron beam 330 passing through the rectangular opening 411 into a desired oblong shape is formed. The electron beam 330 irradiated from a charged particle source 430 and passing through the rectangular opening 411 is deflected by a deflector. The electron beam 330 passes through a part of the variable shaping opening 421 and is irradiated on a target object which is placed on a stage and to which a resist material is coated. The stage continuously moves in a predetermined direction (for example, an X direction) during pattern writing. In this manner, an oblong shape which can pass through both the opening 411 and the variable shaping opening 421 is written in a writing region of a target object 340. A scheme which causes an electron beam to pass through both the opening 411 and the variable shaping opening 421 to form an arbitrary shape is called a variable shaping scheme. In recent years, as one resist frequently used in electron beam, a chemical amplification type resist is known. The chemical amplification type resist is obtained by blending a photo-acid generator in a resist polymer of a resist. An acid generated in the resist by exposure serves as a catalyst to promote a solubilization reaction or an insolubilization reaction of the resist. As resists, a positive type resist in which a charged particle irradiated portion is solubilized by a developing solution to form a hole by development and a negative type resist in which a charged particle irradiated portion is insolubilized to form a hole in an unirradiated portion, exist. As a base resin, resin materials which can be used in the positive type resist and the negative type resist are different from each other. As the positive type resist, a PMMA (polymethylmethacrylate) developed by a mixed solvent of MIBK (methyl isobutyl ketone) and isopropyl alcohol (IPA) is well known. However, recently, an alkaline solubilization resin resist is also used. As a resist containing an alkaline solubilization resin, a phenol resin, a novolac resin, a substituted polystyrene resin, and the like are given. On the other hand, as an example of the negative type resist, a compound which is cross-linked or polymerized by an acid and insolubilized in an alkaline developing solution can be used. More specifically, an alkylether melamine resin, an alkylether benzoguanamine resin, an alkylether urea resin, an alkylether-group-containing phenolic compound, and the like can be given. As a type of a acid generator, a charged particle beam irradiation acid generator (generally known as a photo-acid generator) which releases and generates acid by irradiating a charged particle, or an acid generator which generates an acid by heating is known. As examples of the charged particle beam irradiation acid generator, compounds such as bis-sulfonium diazomethanes, nitrobenzile derivatives, polyhydroxy compounds, aliphatic or aromatic sulfonic esters, an onium salt, sulfonylcarbonyl alkanes, sulfonylcarbonyl diazomethanes, halogen-containing triazine compounds, oxime-sulfonate-based compounds, and phenylsulfonyloxyphthal imides can be used. On the other hand, as a thermal acid generator, sulfonimide is known. The sulfonimide generates an acid in a temperature range of 140° to 150°. The chemical amplification type resist has a problem that negligence before and after exposure makes an optimum exposure to fluctuate. In other words, when a chemical amplification resist is used in mask manufacturing, a line width (CD) after writing of a mask serving as a target object fluctuates. A fluctuation (PED) of the line width (CD) after the writing of the mask may be caused by diffusion of an acid generated by pattern writing. As a method of solving the problem, a pattern width state is recorded up to a writing end in a test mode, and, in actual pattern writing, a shot amount correction value is calculated from a pattern width at a start of pattern writing in the test mode, a resist sensitivity ratio, and a writing prediction time. A technique which adds the shot amount correction value to writing data when the correction is not performed to correct pattern writing accuracy is disclosed in Published Unexamined Japanese Patent Application No. 2006-303361, for example. In addition, a technique in which although not in mask writing, when a wafer to which a chemical amplification type resist is coated is exposed by using an electron beam, an exposure is determined on the basis of a negligence time is disclosed in Published Unexamined Japanese Patent Application No. H9-237745, for example. A technique which corrects a shot size of a beam by an elapsed time from a start of pattern writing is disclosed in Published Unexamined Japanese Patent Application No. 2007-34143, for example. In this case, when a pattern writing process is performed, a process which is not directly related to the pattern writing process, for example, operations in a writing apparatus such as a robot operation related to a conveying operation of a mask serving as a target object for pattern writing, a valve opening/closing operation, and an actuating operation of a vacuum pump may be collaterally performed. However, a track of an electron beam in pattern writing changes by an influence of noise, magnetic field fluctuation, or the like caused by these operations, causing the positions of a beam to be irradiated to fluctuate. This error cannot be allowed with micropatterning in recent years. In this manner, the error disadvantageously influences pattern writing accuracy. For this reason, during the conveying operation as described above, a method of temporarily stopping pattern writing may be performed. However, when pattern writing is performed on a target object such as a mask substrate to which the chemical amplification type resist is coated, the following problem is posed at the temporary stop of the pattern writing. When pattern writing is performed on a target object to which a chemical amplification resist is coated, a writing prediction time required to write a pattern on one target object is calculated in advance, and a correction which changes a exposure dose depending on an elapsed time of pattern writing is performed such that a desired dose is obtained at the end of pattern writing. However, since a timing at which pattern writing is temporarily stopped changes depending on registration contents of a pattern writing job or processing states of the pattern writing job, an actual writing time may be considerably different from a predicted pattern writing time. For this reason, a dose correction function does not effectively work. As a result, pattern writing accuracy may be deteriorated. It is an object of the present invention to provide a pattern writing method which suppresses pattern writing accuracy from being deteriorated. A charged particle beam writing apparatus according to an embodiment of the present invention includes writing a pattern on a first target object by using a charged particle beam in a writing apparatus; and conveying a second target object after having written the pattern on the first target object, wherein even though the second target object is arranged on any one of conveying paths including a carry-out port and a carry-in port of the writing apparatus, a conveying operation for the second target object is not performed during writing the pattern on the first target object. A charged particle beam writing method according to another embodiment of the present invention includes writing, by using a charged particle beam, a pattern on a target object to which a chemical amplification type resist is coated; determining the presence/absence of a request of an item which stops a pattern writing operation for the target object; and determining whether a stop time of the pattern writing operation occurred by performing the item is included in a pattern writing prediction time of the target object, wherein when the stop time of the pattern writing operation is not included in the pattern writing prediction time, the pattern writing operation for the target object is continued without performing the item regardless of the request of the item. In an embodiment, a configuration using an electron beam will be described as an example of a charged particle beam. The charged particle beam is not limited to the electron beam. A beam using other charged particles such as an ion beam may be used. FIG. 1 is a conceptual diagram showing a configuration of a writing apparatus according to Embodiment 1. In FIG. 1, a writing apparatus 100 includes a pattern writing unit 150, a control unit 160, a carry-out/in port (I/F) 120, a load lock chamber 130, a robot chamber 140, a pre-chamber 146, and a vacuum pump 170. The writing apparatus 100 serves as an example of a charged particle beam writing apparatus. The writing apparatus 100 writes a desired pattern on a target object 101. The control unit 160 includes a control computer 110, a storage device 112 such as a memory or a magnetic disk device, and a drive circuit 114. The pattern writing unit 150 has an electron mirror barrel 102 and a pattern writing chamber 103. In the electron mirror barrel 102, an electron gun assembly 201, an illumination lens 202, a first aperture plate 203, a projection lens 204, a deflector 205, a second aperture plate 206, an objective lens 207, and a deflector 208 are arranged. On the X-Y stage 105, the target object 101 is arranged. In the carry-out/in port 120, a conveying robot 122 which conveys the target object 101 is arranged. In the robot chamber 140, a conveying robot 142 which conveys the target object 101 is arranged. The vacuum pump 170 exhausts a gas from the robot chamber 140 through a valve 172. In this manner, the robot chamber 140 is maintained in a vacuum atmosphere. The vacuum pump 170 evacuates gases from the electron mirror barrel 102 and the pattern writing chamber 103 through a valve 174. In this manner, the electron mirror barrel 102 and the pattern writing chamber 103 are maintained in a vacuum atmosphere. At boundaries between the carry-out/in port 120, the load lock chamber 130, the robot chamber 140, and the pattern writing chamber 103, gate valves 132, 134, and 136 are arranged, respectively. As the target object 101, for example, a exposure mask substrate for transferring a pattern to a wafer is included. The mask substrate, for example, includes mask blanks on which any pattern is formed. Data input/output or calculated in the control computer 110 is stored in the storage device 112 in each case. The drive circuit 114 is controlled by the control computer 110. According to the control contents, devices in the pattern writing unit 150, the carry-out/in port 120, the load lock chamber 130, the pre-chamber 146, and the robot chamber 140 are driven. In this case, in FIG. 1, configuration components required to explain Embodiment 1 are described. The writing apparatus 100 may generally include another necessary configuration, as a matter of course. In addition, the conveying robots 122 and 142 may be mechanical mechanisms such as elevator mechanisms or rolling mechanisms. The electron beam 200 emitted from the electron gun assembly 201 serving as an example of an irradiating unit entirely illuminates the first aperture plate 203 having an oblong, for example, rectangular hole by using the illumination lens 202. In this case, the electron beam 200 is shaped into an oblong, for example, rectangle. The electron beam 200 of a first aperture image passing through the first aperture plate 203 is projected on the second aperture plate 206 by the projection lens 204. A position of the first aperture image on the second aperture plate 206 is controlled by the deflector 205 to make it possible to change a beam shape and a beam size. As a result, the electron beam 200 is shaped. The electron beam 200 of the second aperture image passing through the second aperture plate 206 is focused by the objective lens 207 and deflected by the deflector 208. As a result, the beam is irradiated on a desired position of the target object 101 on the X-Y stage 105 which continuously moves. FIG. 2 is an upper conceptual diagram showing a conveying path in the writing apparatus according to Embodiment 1. The target object 101 arranged in the carry-out/in port 120 is conveyed onto a stage in the load lock chamber 130 by the conveying robot 122 after the gate valve 132 is opened. After the gate valve 132 is closed, the gate valve 134 is opened, the target object 101 is conveyed onto a stage in the pre-chamber 146 by the conveying robot 142 through the robot chamber 140. The target object 101 is on standby in the pre-chamber 146, and, thereafter, the gate valve 136 is opened to convey the target object 101 onto the X-Y stage 105 in the pattern writing chamber 103. After the gate valve 136 is closed, a predetermined pattern is written onto the target object 101 on the X-Y stage 105. After the pattern writing is ended, the gate valve 136 is opened, the target object 101 is moved from the X-Y stage 105 in the pattern writing chamber 103 into the robot chamber 140 by the conveying robot 142. After the gate valve 136 is closed, the gate valve 134 is opened, the target object 101 is conveyed onto the stage in the load lock chamber 130 by the conveying robot 142. After the gate valve 134 is closed, the gate valve 132 is opened, the target object 101 is conveyed to the carry-out/in port 120 by the conveying robot 122. In these operations, each time a degree of vacuum in each chamber decreases, the vacuum pump 170 operates to maintain the vacuum state. Alternatively, the valve 172 or the valve 174 is opened or closed, and evacuation is performed by the in-operation vacuum pump 170 to maintain a desired degree of vacuum. When a pattern is written on the target object 101a in the pattern writing chamber 103, a next target object 101b is conveyed into the pre-chamber 146. A second next target object 101c may be on standby in the carry-out/in port 120. In this case, in the middle of pattern writing on the target object 101c in the pattern writing chamber 103, when a conveying operation of the target object 101b on which a pattern is next and subsequently written is performed, as described above, the electron beam 200 is influenced by the conveying operation, and pattern writing accuracy is deteriorated. More specifically, a track of the electron beam 200 in pattern writing is changed by influences of noise, a magnetic field fluctuation, and the like caused by conveying operations such as the operations of the conveying robots 122 and 142, the opening/closing operations of the gate valves 132, 134, and 136, or the actuating operation of the vacuum pump 170, and a position of a beam to be irradiated fluctuates. Therefore, in Embodiment 1, the following countermeasure is performed. In this case, a target object 101 onto which a chemical amplification type resist is coated, is explained below as an example, where the effect of the present invention is more prominent. However, a target object to which a resist except for the chemical amplification type resist may be used. FIG. 3 is a flow chart of a writing method according to Embodiment 1. In step S102, as the pattern writing step, in the pattern writing chamber 103 of the writing apparatus 100, by using the electron beam 200, pattern writing on the target object 101a (first target object) is started. It is assumed that the chemical amplification type resist is coated, or “applied” to the target object 101a. A pattern writing prediction time until the pattern writing operation on the target object 101a is ended is estimated in advance, and a dose (exposure dose) is set on the basis of the pattern writing prediction time. In place of the chemical amplification type resist, a resist the sensitivity of which changes with time is preferably used. FIG. 4 is a diagram for explaining a method of correcting a exposure dose in Embodiment 1. A dose D(t) used in actual pattern writing can be defined as follows by using a time t elapsed from the start of pattern writing, a pattern writing prediction time Te, a dose correction coefficient δD, and a reference dose D0. The equation can be expressed by a linear function of, for example, D(t)=D0−δD(Te−t). A pattern size x(t) can be similarly defined as follows by using the time t elapsed from the start of pattern writing, the pattern writing prediction time Te, a pattern size correction coefficient δx, and a reference size x0. The equation can be expressed by a linear function of, for example, x(t)=x0−δx(Te−t). In S104, as the first determining step, a first determining process function in the control computer 110 determines the presence/absence of a request event (item) which stops a pattern writing operation. This event corresponds to a conveying operation. The conveying operation includes at least one of the operations of the conveying robots 122 and 142, the opening/closing operations of the gate valves 132, 134, and 136, and an actuating operation of the vacuum pump 170. When the event is requested as a result of the determination, the operation shifts to S106. When the event is not requested, the operation shifts to S116. When the first determining step is set, the presence/absence of the event which stops the writing operation can be comprehended. In S106, as the second determining step, a second determining process function in the control computer 110 determines whether a time (stop time) for stopping a pattern writing operation occurred by performing the event is included in the pattern writing prediction time Te of the target object 101a. When the time is included in the pattern writing prediction time Te as a result of the determination, the operation shifts to S110. When the time is not included in the pattern writing prediction time Te, the operation shifts to S108. When the second determining step is set, it can be comprehended whether the event is included in the pattern writing prediction time Te. In S108, when the time for stopping the pattern writing operation is not included in the pattern writing prediction time Te, the pattern writing process function in the control computer 110 continues the pattern writing operation without performing the event regardless of the request of the event. More specifically, even though the target object 101b (second target object) on which a pattern is written next is arranged in the carry-out/in port 120 of the writing apparatus 100, the target object 101b is on standby with the conveying operation of the target object 101b not performed during writing a pattern on the target object 101a. In this manner, the electron beam 200 can be prevented from being influenced by the conveying operation during the pattern writing. In particular, when a pattern is written on the target object 101a to which the chemical amplification type resist is coated, the pattern writing prediction time Te is advantageously suppressed from shifting. In S110, when the time for stopping the pattern writing operation is included in the pattern writing prediction time Te, the pattern writing process function in the control computer 110 temporarily stops the pattern writing operation while the event is being performed. In S112, the event processing function in the control computer 110 executes a requested event. Periodical apparatus diagnosis such as current density measurement is given as an example. As a period for the measurement, for example, the measurement is performed for 1 minute once 15 minutes. The times for these periodical events are preferably included in the pattern writing prediction time Te in advance. More specifically, an increase in time by the current density measurement is added to the prediction time obtained by only the pattern writing operation. When the initial pattern writing prediction time (execution time) is 10 hours (600 minutes), it is assumed that the total pattern writing prediction time of 640 minutes is obtained by adding the increase in time by the current density measurement. Therefore, the pattern writing prediction time Te is set at 640 minutes. In this manner, even in pattern writing in which correction is performed by an elapsed time, with respect to periodically performed processes, an increase in the pattern writing time is considered by using an execution time calculated in advance and a set event period, so that highly accurate corrected pattern writing can be realized. Alternatively, when a time required for even an event occasionally occurring is known in advance, and when an increase in pattern writing time by the stop of pattern writing by the event falls within a preset allowed time, the pattern writing operation may be temporarily stopped to execute the event. For example, as shown in FIG. 4, when an increase/decrease in time corresponding to an allowed size error ±Δx is ±Δt, a pattern writing operation may be temporarily stopped even for an event occurring occasionally, if the process of that event ends within the period Δt. For example, about ±15 minutes can be allowed for pattern writing for 10 hours. In this manner, when the pattern writing prediction time Te has a predetermined margin, the pattern writing operation is preferably stopped while the event is executed when a time for stopping the pattern writing falls within a predetermined margin. In S114, a pattern writing process function in the control computer 110 restarts a pattern writing operation after an event is ended. The operation shifts to S116. In S116, as the third determining step, a third determining process function in the control computer 110 determines whether continuation of the pattern writing operation is necessary. When the pattern writing is not ended, the operation returns to S104. When the continuation of the pattern writing operation is not necessary, the operation shifts to S118. In S118, when the pattern writing process function in the control computer 110 ends pattern writing on the target object 101a, the pattern writing operation is ended. In S120, as the fourth determining step, the fourth determining process function in the control computer 110 determines whether a required event is executed. When the event is executed, the operation is ended. Still in a standby state, the operation shifts to S122. In S122, an event processing function in the control computer 110 executes a requested event. More specifically, even though the target object 101b is arranged at the carry-out/in port 120 of the writing apparatus 100, a conveying operation of the target object 101b is not performed while a pattern is written on the target object 101a, and the target object 101b is conveyed after the pattern writing on the target object 101a is ended. After the event is executed, the flow is ended. As described above, according to the embodiment, an influence of a conveying operation can be excluded. In particular, when a pattern is written on a target object to which a chemical amplification type resist is coated, a pattern writing prediction time can be suppressed from shifting. Therefore, pattern writing accuracy can be suppressed from being deteriorated. In the above explanation, the “units” or the “steps” described above can be programs operated by a computer. The units or the steps may be executed by not only a program serving as software but also a combination between hardware and software or a combination between hardware and firmware. When the units or the steps are constituted by programs, the programs can be recorded on a readable recording medium such as a magnetic disk device, a magnetic tape device, an FD, a CD, a DVD, an MO, or a ROM. For example, the programs are stored in the storage device 112. At least one of these recording media may be connected to the control computer 110. The recording medium may be arranged in the control computer 110. The embodiment is described with reference to concrete examples. However, the present invention is not limited to these concrete examples. Parts such as an apparatus configuration and a control method which are not directly necessary for the explanation of the present invention are omitted. However, a necessary apparatus configuration and a necessary control method can be arbitrarily selected and used. For example, a description of a control unit configuration for controlling the writing apparatus 100 is omitted. However, a necessary control unit configuration can be arbitrarily selected and used, as a matter of course. Furthermore, all charged particle beam writing methods and apparatuses which can be arbitrary changed in design by a person skilled in the art are included in the scope of the present invention. Additional advantages and modification will readily occur to those skilled in the art. Therefore, the invention in its broader aspects is not limited to the specific details and representative embodiments shown and described herein. Accordingly, various modifications may be made without departing from the spirit or scope of the general inventive concept as defined by the appended claims and their equivalents.
052987590	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS The photo-cracker cell of the invention is a device for producing a monomeric (atomic) beam of Group V or Group VI atoms by photodissociation of the effluent from a conventional thermal effusion cell in MBE, a thermal cracker in gas-source MBE, or parent gas molecules from an injector in MOMBE. The atomic beam thus produced can be used for the growth or doping of semiconductor thin films. The availability of an atomic beam offers distinct advantages over conventional sources in terms of control of the composition and doping of layers grown by MBE, gas-source MBE, and MOMBE. Currently, MBE gas and solid sources utilize thermal "cracking" in an attempt to dissociate the tetramer and dimer components that are produced by simple molecular effusion sources. However, the generation of monomeric beams of Group V and VI elements by thermal dissociation is very inefficient, and the monomer fraction from such sources is extremely low. In the present invention, UV light is used to photodissociate the output beam from a thermal cell, thereby achieving a very high fractional content of monomeric species in the beam. As contemplated herein, the Group V species typically comprises arsenic (As) and antimony (Sb), while the Group VI species comprises tellurium (Te) and selenium (Se). However, it will be readily appreciated that the teachings of the present invention are not to be construed as limited to these particular species. The purpose of the photo-cracker cell of the invention is to generate pure atomic beams, that is, beams of monomeric species, of Group V and VI elements in order to improve the control of composition and doping of semiconductor layers by MBE, gas-source MBE, and MOMBE. This improvement results from the increased sticking coefficient of monomeric species on a semiconductor surface compared to that of dimer or tetramer clusters, and the ability to monitor and control the flux of an atomic beam. The low efficiency of presently-employed thermal cracking of Group V and VI clusters in MBE, gas-source MBE, and MOMBE leads to a large fraction of dimers and tetramers in the beam. This limits the precision of composition and doping control because of the low sticking coefficients of the molecular species and the effects of co-adsorption on the incorporation rate of atoms into the growing layer. An example is described for the growth of GaAs.sub.x Sb.sub.1-x using a monomer source for the antimony flux (which has near-unity sticking coefficient). In this application, an antimony flux is established which produces the desired 1-x value with respect to the Ga flux. The As flux is then allowed to flood the surface in excess to provide sufficient Group V species to react and combine with the remaining gallium, thus producing the desired alloy. Since the sticking coefficient of atomic Sb is high, virtually all of the Sb flux contributes to layer growth. The As over-pressure provides the remaining Group V species for utilization of the Ga flux, and the excess is re-evaporated (as occurs in conventional GaAs growth, in which an excess As flux is also employed). The applications for an atomic cracker cell extend from (a) III-V material growth, where the Group V source has a direct influence on growth rate and alloy composition, to (b) Group IV material growth, where Group V elements are used as n-type dopants, and to (c) II-VI growth, where Group VI elements control the composition and Group V elements are dopants. Thus, the potential applications of a monomer source are very diverse. The basic embodiment of the photo-cracker cell of the present invention is depicted in FIGS. 1-2. The photo-cracker cell 10 utilizes ultraviolet (UV) light from a flashlamp 12 to photodissociate molecular beams 14 comprised of Group V or VI clusters. The photo-cracker cell is positioned in the output stream of a source cell, indicated at 16. By "source cell" is meant a thermal effusion cell in MBE, a thermal cracker cell in gas-source MBE, or a gas injector in MOMBE. In the embodiment shown in FIGS. 1-2, the flashlamp 12 is placed along one focus of an elliptically-shaped, reflective cavity 18, with the molecular beam path 14 lying along the other focus. The relative placement of the flashlamp 12 and the molecular beam path 14 in the elliptically-shaped, reflective cavity 18 is depicted in FIG. 2. In this configuration, most of the light emanating from the lamp 12 is collected and focused onto the beam 14 of molecules, and thus the highest photodissociation efficiency can be achieved. One concern is that the flashlamp 12 might gradually become coated with the material from the beam 14 due to the random transverse component of particle velocity following photodissociation. Coating of the lamp 12 due to this effect can be prevented or at least minimized, with little penalty in dissociation efficiency, by interposing a curved, reflective shield 20 between the lamp and the beam path to block the line-of-sight path between the lamp and beam 14. This shield 20 is shown in the perspective drawing of FIG. 1 and the top plan view of FIG. 2. In the embodiment described, the photo-cracker cell is placed at the end of a thermal effusion cell in MBE, at the end of a thermal cracker cell in gas-source MBE, or at the gas injector of MOMBE. Some modification to the existing apparatus may be required to accommodate the photo-cracker cell in this embodiment; however, in this configuration, the source-to-sample distance is not substantially increased. FIGS. 3-4 depict a mechanical configuration for the photo-cracker cell of the invention. In this configuration, the flashlamp 12 is housed inside an ultrahigh vacuum (UHV)-compatible quartz tube 22 which is open to air on one end. The quartz tube 22 extends into a vacuum chamber housing 24, through which the beam 14 also passes. This arrangement provides for easy replacement of the lamp 12 without opening the vacuum chamber housing 24 and also facilitates cooling of the lamp with either water or air. The quartz tube 22 is secured to the vacuum chamber 24 by removable means 26. As shown in FIG. 4, the photo-cracker cell 10 can be built with conventional vacuum flanges 28, 30 near the entrance 32 and exit 34, respectively, thereof, so that it can be conveniently added to any existing MBE, gas-source MBE, or MOMBE system by simply interposing the cell between the growth chamber (shown generally at 36) and an existing source cell 16 (effusion cell, thermal cracker cell, or gas injector). In such an event, the change in geometry may cause a reduction in Group V or Group VI beam flus. However, even if the flux is reduced, the greater efficiency of incorporation obtained using the substantially monomeric beam more than compensates for this tradeoff. Alternatively, exit portion 34 and flange 30 may be eliminated and replaced by a larger flange 35 at the base of the cell 10, as shown in FIG. 4, for incorporation into the growth chamber, the flange 35 being bolted onto the wall of the growth chamber, and thereby permitting a reduction in the distance between the source and the substrate from that of the configuration immediately described above. In an alternate embodiment, depicted in FIG. 5, a UV laser source 38 may be used in place of the flashlamp 12 of FIGS. 1-4, with little modification to the cell design. The UV laser source 38 is positioned outside the vacuum chamber 24, with the UV radiation 40 introduced into the vacuum chamber through an opening 26a at means 26 and distributed into the chamber by a plurality of angled, partially reflecting mirrors 42 (of appropriate reflectivity to ensure substantially uniform irradiation) positioned serially along the length of the quartz tube 22. The angle of the mirrors is off of the line, or axis, defined by the quartz tube 22 and the molecular beam path 14. For example, the mirrors may be 45.degree. off-axis. As above, the UV radiation 40 is reflected by the elliptical cavity 18 to illuminate a substantial portion of the length of the beam 14. The UV radiation employed is dependent on the molecular species being cracked. For example, cracking As.sub.4 (tetramer) to monomeric (atomic) As requires a different wavelength than cracking As.sub.2 (dimer) to monomeric As. Thus, while a single wavelength source, such as UV laser 28 may be employed in the practice of the invention, preferably, UV lamp 12, which provides a broader wavelength range, is employed. Such a UV lamp is advantageously a high intensity discharge lamp, and may be continuous or pulsed to provide a sufficiently high photon output to photodissociate substantially all the dimer and tetramer species. If the UV light source is pulsed, the pulse length must be long enough and the intensity of the photon flux must be high enough to photodissociate substantially all the molecules (i.e., dimers and tetramers) in the irradiated portion of the molecular beam. Further, the repetition rate, or time between pulses, must be short compared to the transit time of the Group V or Group VI species along the beam path 14. As used herein, the UV wavelength range applicable in the practice of the present invention generally ranges from about 180 to 400 nm. The highest efficiency of photo-cracking is obtained by providing an elliptical focussing arrangement, as described above, and accordingly, this arrangement is preferred. However, other focussing configurations and even non-focussing configurations may alternatively be employed in accordance with the teachings of this invention. The beam of Group V or Group VI species that is produced in accordance with the invention is substantially monomeric. By this is meant that whereas thermal cracking produces a beam that, at best, contains about a small fraction of monomeric species, the present invention, employing photo-cracking (UV light), is expected to produce a beam that contains at least about a ten-fold increase in monomeric species. Thus, there has been disclosed a method and apparatus for the production of substantially monomeric Group V or Group VI metal species. It will be readily apparent to those skilled in this art that various changes and modifications of an obvious nature may be made, and all such changes and modifications are considered to fall within the scope of the invention as defined by the appended claims.
062366998	description	DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 is a diagrammatic illustration of the CEA Rod Position System in accordance with the present invention, for a pressurized water nuclear reactor. The reactor is controlled by CEA's which are actuated by drive mechanisms 22 which, in turn, are controlled by a Control Element Drive Mechanism Control System (CEDMCS) 50 (or equivalent rod control system). During operation of a pressurized water reactor, the coolant is circulated through the reactor core, contained within the nuclear rector vessel 10, which extracts heat from the core and heats the coolant. This heated primary coolant is then passed through a steam generator 12 where it exchanges its heat with a secondary coolant that circulates through the secondary side of the steam generator 12. After transferring its heat to the secondary coolant, the primary coolant is then recirculated by reactor coolant pumps 19 back to the reactor 10. The secondary coolant, which is ordinary water, is heated from its normal liquid phase to vapor phase as a consequence of the heat transfer from the primary coolant which occurs in the steam generators 12. It is then passed to the plant turbine 14 which converts the heat energy of the vaporous phase into mechanical energy. The secondary coolant is then condensed and is recirculated to the steam generator 12 by means of steam generator feedwater pumps 16. A digital computer 54 receives certain plant input signals and then processes them to perform the requisite on-line monitoring of CEA Regulating and Part Strength rod groups relative to the Limiting Conditions for Operation (LCO) for insertion between the Long Term Steady State Insertion Limit and the Transient Insertion Limit. In order to properly perform this function, the digital computer 54 requires the CEA rod positions for the Regulating and Part Strength rods, indications if a reactor power cutback condition is present, and the values of certain plant parameters. The CEA rod positions are obtained from the Control Element Drive Mechanism Control System (CEDMCS) 50 (or equivalent rod control system). The CEDMCS determines rod positions for each Control Element Assembly by keeping an aggregate count of the number of "up" and "down" pulses which are generated by CEDMCS whenever a request is sent to move the rods up or down by one position (e.g.--0.75 inches or 1.9 cm of vertical displacement) to the control rod control mechanisms 22. These signals are generated within CEDMCS by an Automatic CEA Timing Module (ACTM) which outputs "up" and "down" pulses of 300 to 500 millisecond duration. The "up" and "down" pulses are accumulated within CEDMCS via pulse counters for each of the rods. The rod positions are determined within CEDMCS by taking the difference in aggregate pulse counts between the "up" and "down" pulses and multiplying this difference by the vertical displacement due to a single pulse command. EQU Rod Position=(S{up pulses}-S{down pulses}).times.0.75 inch Pulse counters are reset to zero whenever a rod is fully inserted into the core and encounters a "bottom contact" switch. The individual CEA rod positions, as determined via CEDMCS pulse counting, are transmitted from CEDMCS 50 to the digital computer 54 via a data link. These values are then stored in the plant computer data base 60 where they are accessible by the CEA Rod Position Program 56. Alternately, rod positions can be directly determined via a control rod position detector 24 which is common to pressurized water nuclear reactors. Indication of the initiation of an accelerated power reduction, known as Reactor Power Cutback (RPC), is determined by the RPC Monitor 52. The RPC Monitor notes whenever an event is present which requires an accelerated power reduction (such as the failure of a steam generator feedwater pump 16) and sends a signal to the digital computer 54 indicating that a RPC event is present. The status of the RPC condition (RPC event present or not present) is stored in the plant computer data base 60 where it is accessible by the CEA Rod Position Program 56. Certain plant parameter values are required by the digital computer 54 in order for the COLSS program 58 to calculate the reactor power level. These values are obtained from the plant sensors as follows: primary coolant flow rate 32, primary coolant pressure 34, cold leg temperature (T cold) 36, hot leg temperature (T hot) 38, feedwater temperature 40, feedwater flow 42, steam flow 44, and steam pressure 46. These values are then stored in the plant computer data base 60 where they are accessible by the COLSS program 58 (or equivalent). COLSS 58 utilizes this data to determine plant power by determining the net energy leaving a control volume taken out of the reactor (primary calorimetric method) and by performing an energy balance based on the plant secondary system (secondary calorimetric method). Alternately, plant power may be directly determined via the neutron flux detectors 30 which are common to pressurized water nuclear reactors. The COLSS program 58 computes plant power and stores this value along with the COLSS program status (operating mode, off-line mode, test mode) into the computer data base 60 where it can be accessed by the CEA Rod Position Program 56. The CEA Rod Position Program 56 determines if the restrictions imposed upon CEA rod group insertions between the Long Term Steady State Insertion Limit and the Transient Insertion Limit (expressed in terms of hours and EFPD exposure) are maintained. This program acquires individual rod positions (as determined via CEDMCS 50) from the data base 60, acquires the LCO limits for CEA insertion from the data base 60, acquires RPC status (as determined via the RPC monitor 52) from the data base 60, and acquires the value of plant power and COLSS status (as determined via COLSS 58) from the data base 60. A magnetic data storage disk 62 (or equivalent long term data storage device) is used to store CEA exposure files and CEA historic records which are utilized by the CEA Rod Position Program 56. A keyboard 64 accepts operator inputs and a CRT,66 displays output data and alarm annunciation messages. (Alternate data entry and display devices, such as touch screens and LCD flat panel displays, may also be utilized in place of the CRT and keyboard.) Positions for the Regulating Groups and Part Strength Groups, as used in the CEA Rod Position Program 56, are determined by the "Middle Group Average" method. This method eliminates the highest and lowest CEA's from the averaging calculation so that the group average position is based on the "middle" CEA positions, and is not skewed by an unusually high or low individual CEA element position. This method is illustrated below for the case of Regulating Group 1 which consists of "n" CEA elements whose positions are stored in array RG1(J). The calculation proceeds as follows: EQU P1 min=Min (RG1(1), RG1(2), . . . RG1(n) EQU P1 max=Max (RG1(1), RG1(2), . . . RG1(n) EQU P1 sum=S{ RG1(j)}, j=1 to n EQU RG1--Position=(P1 sum-P1 min-P1 max)/(n-2) Where: "P1 min" contains the value of the lowest CEA position for Reg Group 1; "P1 max" contains the value of the highest CEA position for Reg Group 1; "P1 sum" contains the sum of all CEA rod positions for Reg Group 1; "n" contains the number of CEA elements in Reg Group 1; "RG1--Position" is the average position of Regulating Group 1 as determined by the "Middle Group Average" method. These calculations may be performed within the data base using "composed data base points" (in which the aforementioned calculations are performed directly within the data base), or alternately, can be performed as a separate CEA group position module associated with the CEA Rod Position Program 56. FIG. 2 provides an overview of the software structure for the CEA Rod Position Program 56 for the key operating functions. Box 150 is the Program Executive which controls and schedules the execution of the other program modules which comprise the CEA Rod Position Program (The specific instructions contained therein are dependent upon the particular software operating system which is employed by the digital computer). Box 151 performs a calculation of Effective Full Power periods and then sequentially calls the next four modules (boxes 152 to 155) in the chain. Box 152 performs an update of the CEA exposure accumulation history for all Regulating and Part Strength groups, box 153 performs a determination of the current value of CEA exposure for all Regulating and Part Strength groups, box 154 scales the exposure accumulation data so it conforms to the pixel display constraints of the display CRT, and finally box 155 draws the graphical output in either "rolling wheel" or "sliding bar" display formats. Box 156 supports the recall of historic CEA exposure accumulation records. Box 157 contains the main Predictor Module which predicts when sufficient CEA exposure margin will be regained to allow a planned CEA rod maneuver without violating the LCO's. Boxes 158 to 161 contain software modules which support additional calculations that are required for the Prediction Function. Box 158 performs margin prediction calculations, box 159 performs a mapping of CEA exposure history from array "P(J)" (which contains the current CEA exposure data) into array "PP(J)" (which is utilized for exposure history prediction calculations), box 160 updates the CEA exposure margin for the predictor array "PP(J)", and box 161 calculates CEA exposure margin using the predictor array "PP(J)". Box 162 contains the main update module which is used to update time dependent CEA exposure data in the event of a computer outage of the CEA Rod Position System. After execution of box 162, then box 163 is called. Box 163 performs additional calculations on the update intervals (by shifting elements of array "P(J)" up by 1 position) that are required to support the Update Function. The detailed functional operation of these software modules is subsequently described herein. Prior to describing the requisite logic associated with this system, an understanding of the data storage structure is a helpful prerequisite in order to follow the subsequent algorithmic functions which operate directly upon the stored data. The following descriptions are provided for the case utilizing data storage intervals based upon Effective Full Power (EFP) criteria (EFP hour intervals) rather than standard time intervals (hours) as these are the more complex data recording intervals to maintain. (For criteria based on standard time intervals (hours), the array elements correspond to "hours" instead of "EFP hours" and there is no correspond calculation required to determine "EFP hourly intervals".) FIG. 3 illustrates a generic data array structure for maintaining the CEA exposure records. There are multiple arrays to track each of the CEA exposure criteria (e.g.--5 EFPD per 30 EFPD interval for Regulating groups, 14 EFPD per 365 EFPD interval for Regulating groups, 7 EFPD per 30 EFPD interval for Part Strength groups, etc.). Each of the Regulating and Part Strength groups has a corresponding set of such arrays to track the requisite CEA exposures. Each of these arrays consists of N elements. The number of elements (N) is dependent upon the CEA exposure which is being maintained (e.g.--5 EFPD per 30EFPD interval for Regulating groups, etc.). Each element in an array corresponds to a fixed Effective Full Power (EFP) interval of exposure, with element 1 corresponding to the most recent EFP interval and element N corresponding to the oldest EFP interval. As an example, for the "5 EFPD per 30 EFPD interval" criteria for Regulating groups, and utilizing an EFP data recording interval of 1 EFP-hour, a total of 12430=720 array elements would be required to contain 30 days worth of EFP exposure, recorded at a resolution of 1 EFP-hourly intervals. For this array, element N=720 would correspond to the oldest data (720 EFP-hours old) while element N=1 would correspond to the most recent data (1 EFP-hour old)). EFP-hourly intervals are computer based upon plant power operating level and time duration (for example--with a plant power level of 0.5 EFP, a two hour time interval would be required to obtain a 1 EFP-hour interval), The contents of each array element indicate what the CEA exposure accumulation was for the CEA group during the EFP-hourly interval that the array element corresponds to. If CEA exposure accumulation occurred during an EFP interval, then the total exposure accumulation which occurred during that interval is stored as the array element value; otherwise a "0" is stored (for example, if a Regulating group was inserted between the Long Term Steady State Insertion Limit and Transient Insertion Limit for 1/2+L the time during the EFP hourly period that occurred 4 EFP hours ago, then the contents of array element N=4 would be "0.5 EFP hour", i.e., P(4)=0.5). FIG. 4 provides a flowchart representation of the EFP COMPUTATIONAL MODULE which is utilized to determine Effective Full Power (EFP) hourly intervals and to determine the EFP exposure which has occurred during these intervals. This module runs periodically, at 30 second intervals, under the direction of the PROGRAM EXECUTIVE (other time intervals may be utilized if increased resolution in computing the 1 EFP-hour interval is desired). With reference to FIG. 4, the EFP COMPUTATIONAL MODULE initially reads the current status of COLSS and Reactor Power Cutback via box 200. Then, via box 201, it determines if COLSS is out of service (determined via the COLSS module, box 58 of FIG. 1) or if a Reactor Power Cutback condition exists (determined via the RPC Monitor, box 52 of FIG. 1). If either of these conditions is present, the module bypasses the computation of an EFP interval and waits until the next 30 second scheduled execution interval. If COLSS is in service and there is no current Reactor Power Cutback condition, the present value of plant reactor power is read at box 202. Then, via box 203, this is divided (normalized) by the sampling interval for this variable (1/120+L hour which corresponds to the 30 second scheduled program execution rate) and then summed with variable "Phr" which is used to accumulate the 30 second "normalized" values of plant reactor power ("Phr" is set to zero as an initiation task upon program bootup and is reset to zero after each 1 EFP-hour period is calculated, by box 218). Boxes 204 to 206 next determine if the accumulated value of "Phr" is equal to or greater than a 1 EFP-hour interval and if true sets variable STOP to 1, or if not true sets variable STOP to 0. Then boxes 207 to 211 determine if the positions of each of the Regulating Groups and Part Strength groups (a total of `Kreg` such positions which are contained in array "Pgroup(k)") lie between the Long Term Steady State Insertion Limit (L1) and the Transient Insertion Limit (L2). For such groups, box 212 then updates the corresponding EFP exposure by accumulating the current EFP normalized value for this interval in array "Pnew(k)" (Array "Pnew(k)" is set to zero as an initiation task upon program bootup and is reset to zero after each 1 EFP-hour period is calculated, by box 218). Box 213 next determines if a 1 EFP-hour interval has occurred (this occurs when STOP=1). If this is true, box 214 performs a call to the CEA EXPOSURE ACCUMULATION HISTORY MODULE to update the CEA exposure history, box 215 performs a call to the CEA EXPOSURE CALCULATION MODULE to update the current value of CEA exposure, box 216 performs a call to the SCALING MODULE to scale the graphical outputs to fit within the pixel constraints of the CRT screen, box 217 performs a call to the DRAWING MODULE to draw the graphical display on the CRT, and then box 218 resets variable "Phr" and the elements of array "Pnew(k)" to zero for use during the next 1 EFP-hour interval calculation. The program then terminates. If box 213 determines that 1 a EFP-hour period has not yet occurred, then the calculation for this 30 second period terminates. In either case, after the program terminates the PROGRAM EXECUTIVE schedules this module for execution again during the next periodically scheduled 30 second interval (box 219). FIG. 5 depicts the functional logic for the CEA EXPOSURE ACCUMULATION HISTORY MODULE. This module is called by the EFP CALCULATION MODULE if condition STOP=1 is true. This modules updates the CEA exposure for the CEA Exposure Data Arrays. When called, this module shifts up by one position each element in the CEA Exposure Array (for each of the Program arrays). Thus, the latest CEA exposure which occurred during the previous 1 EFP-hour interval is moved into the first array position, the CEA exposure from the first array position is moved into the second array position, etc. until all CEA exposure data has been shifted up 1 EFP-hour interval in the array. The value of CEA exposure from the last array element "N" (which represents the oldest CEA exposure data) is removed from the array since it is now beyond the LCO EFP duration criteria. This value is stored in the archival records file (where it can be accessed as historical data in conjunction with the HISTORICAL DATA PLAYBACK MODULE which is discussed further below). This process of shifting the contents of each array element up by one position is illustrated in FIG. 6. It is this process whereby a "contiguous monitoring interval" is maintained. For simplicity, the logic for the CEA EXPOSURE ACCUMULATION HISTORY MODULE is illustrated for the case of a single program array (Each of the program arrays, which correspond to the CEA Exposure criteria for each of the Regulating and Part Strength Groups, is similarly operated upon). Referring again to FIG. 5, box 250 first stores the oldest value of "Pn" (from array element N) to the CEA exposure archival file which is contained on disk. The value of "Pn" is saved along with a time stamp that notes the year, date and time that the point was recorded. Box 251 then obtains the CEA Exposure Accumulation History file from the digital computer disk (item 62 on FIG. 1). Boxes 252 to 255 then shift up the contents of the data array elements, beginning from the last array position (that is, first the contents of array element "N131" is shifted into array position "N", then the contents of array element "N132" is shifted into array position "N131", etc.) until array element 1 (the last shift performed by boxes 252 to 255 is from array position 1 to array position 2). After the contents of array position 1 is shifted into position 2, box 256 inserts the value of "Pnew" (the most recent calculated value of CEA Exposure as determined via the EFP Calculation Module) into array position 1. A time stamp is also saved which notes the year, date and hour in which the value of Pnew was determined (this time stamp is utilized when recalling archived historical CEA exposure records). Box 257 then saves the updated CEA Exposure History file to disk storage (via the digital computer disk, item 62 on FIG. 1). The process is repeated until all program arrays are similarly operated upon. For increased computational efficiency, the actual computer implementation of the above process may utilize "circular data storage buffers" for the CEA Exposure Arrays. The shift of positions would then occur by overwriting the oldest CEA exposure value with the newest value and then incremehting software "pointers" which indicate the array starting position (array element=1) and the array ending position (array element=N) within the circular data storage buffer. Thus, the shifting up of the each array element by one position is accomplished with a minimum set of software steps which reduces the computational impact on computer processing resources. The actual logic which would be utilized with circular data storage buffers is dependent upon the chosen hardware/software and is therefore not depicted here. FIG. 7 depicts the functional logic for the CEA EXPOSURE CALCULATION MODULE. This module is called by the EFP CALCULATION MODULE if condition STOP=1 is true. This module updates the CEA exposure for each of the CEA Exposure Data Arrays. For simplicity, the logic for this module is illustrated for the case of a single program array--however, all such aforementioned data arrays are similarly processed. Boxes 300 to 304 calculate the current value of CEA exposure by summing the contents of the CEA Exposure Data Array ("P(I)") in which each array element contains the value of CEA exposure for a given 1 EFP-hour interval. The total accumulated CEA exposure is then stored in variable "EXPOSURE" via box 305. The module then determines the CEA exposure margin ("MARGIN") in box 306 by calculating the difference between the Exposure Limit (such as 5 EFP days) which is stored in variable "LIMIT" and the current value of CEA exposure which is stored in variable "EXPOSURE". Boxes 307 to 309 next determine if there is positive margin (MARGIN>0) or negative margin (MARGIN<0). If the CEA exposure margin is negative (MARGIN<0) then the Alarm Flag is set to one (1) and an alarm is annunciated, alerting the operator that a CEA exposure technical specification has been violated. If the CEA exposure margin is positive (MARGIN>0) then the "Alarm Flag" is set to zero and the CEA exposure margin is further tested by boxes 310 to 312 to determine if the remaining CEA exposure margin ("MARGIN") is less than the pre violation warning limit ("Lwarn"). If the remaining CEA exposure margin ("MARGIN") is less than the pre violation limit (MARGIN<Lwarn), then "Warning Margin Flag" is set to one (1) and a "pre violation CEA exposure" alarm is annunciated, alerting the operator that he is approaching the CEA exposure LCO. If the remaining CEA exposure margin ("MARGIN") is greater than the pre violation limit (MARGIN>Lwarn), then "Warning Margin Flag" is set to zero (0) and no alarm annunciation occurs. The logic for the SCALING MODULE is provided in FIG. 8. For simplicity, the logic for this module is illustrated for the case of a,single program array--however, all data arrays associated with the CEA Rod Position System are similarly processed. This module is called by the EFP CALCULATION MODULE if condition STOP=1 is true. The SCALING MODULE performs a scaling of the CEA Exposure Accumulation Array Elements so that they may be pictorially represented on the CRT (item 66 on FIG. 1) in "rolling wheel" or "sliding bar" formats. The scaling is necessary in order to accommodate the CEA exposure accumulation array information within the pixel constraints imposed by the CRT. The SCALING MODULE examines an interval of CEA exposure accumulation data (such as every 4 consecutive EFP-hourly periods) as stored in the CEA Exposure Accumulation Array Elements (4 consecutive array elements) and determines if any of the array elements within that interval indicate that a CEA exposure accumulation has occurred. If there is any CEA exposure accumulation for the examined interval, the SCALING MODULE then updates a corresponding array (CEA Scaling Array) which is used to drive the output graphics on the CRT. The CEA Scaling Array consists of elements that correspond to each examined interval (such as 4 consecutive EFP-hourly periods) from the CEA Exposure Accumulation Array Elements (that is, 4 consecutive array elements from the CEA Exposure Accumulation Array are mapped into a single array element in the CEA Scaling Array). For cases in which there has been CEA exposure during the examined interval, the SCALING MODULE updates the corresponding array element in the CEA Scaling Array with a one (1), elsewise with a zero (0). The CEA Scaling Array is subsequently utilized by the drawing module to draw either a solid or blank picture segment (depending on the store value in the CEA Scaling Array element--either "1" or "0") for the "rolling wheel" or "sliding bar" output display formats. FIG. 9 illustrates the correspondence between the CEA Exposure Accumulation Array and the CEA Scaling Array. With reference to FIG. 8, boxes 325 and 326 initialize the computation elements for this module. Variable "Z" is set to "N/4" where "N" corresponds to the number of elements in the CEA Exposure Accumulation Array. In this particular case, the SCALING MODULE will scan intervals corresponding to 4 EFP-hours, which corresponds to four consecutive array elements in the CEA Exposure Accumulation Array. Boxes 327 and 328 are used to determine when all such 4 EFP-hour intervals in the CEA Exposure Accumulation Array have been examined (since there are "N" array elements in the CEA Exposure Accumulation Array, then there are Z=N/4 such intervals). Boxes 329 to 332 are used to examine four consecutive array elements in the CEA Exposure Accumulation Array (which corresponds to an interval of 4 EFP-hours). Box 332 determines if any of the four consecutive array elements in the CEA Exposure Accumulation Array contain any CEA exposure accumulation. It performs this determination by examining the contents of each array element for a non-zero value of CEA exposure accumulation (P(4*(J-1)+I)>0 ). If any four consecutive array elements in the CEA Exposure Accumulation Array contain a non zero value, then the corresponding element S(J) of the CEA Scaling Array is updated with a one (1) via box 333, elsewise box 334 updates element S(J) with a zero (0). When box 328 determines that all of the elements of the CEA Exposure Accumulation Array have been examined it then resets variables "J" and "I" via boxes 336 and 337 and sets variable "Kend" to the value of variable "J" via box 335. Variable "Kend" is subsequently used by the DRAWING MODULE. This process of assigning values to array S(J) based on examining the contents of four consecutive elements of array P(N) is illustrated in FIG. 9. Since the CRT has limited pixel resolution relative to the data which is stored in the CEA Exposure Accumulation Array (in this case CEA exposure will be displayed with a granularity of 4 EFP hour intervals), the pictorial displays will have greater "granularity" than the numeric data which is output on the display pages. However, the resolution is still considered sufficient to indicate, pictorially, the relative periods in which CEA exposure accumulation occurred. The numeric data, as output via the normal displays, will always contain the exact values of CEA exposure and the PREDICTOR MODULE will always output when CEA exposure margin will be regained; with a time resolution to the nearest hour. The logic for the DRAWING MODULE is provided in FIG. 10. For simplicity, the logic for this module is illustrated for the case of a single program array--however, all data arrays associated with the CEA Rod Position System are similarly processed. This module is called by the EFP CALCULATION MODULE if condition STOP=1 is true. The DRAWING MODULE provides the graphical outputs for the "rolling wheel" and "sliding bar" display formats. The module functions are described in generalized functional terms as the actual draw commands are dependent upon the specific graphics drawing package which is utilized. Referring to FIG. 10, box 350 determines the requested display format (either "rolling wheel" or "sliding bar" dependent upon the last drawing format selection as made by the operator). Box 351 queue's the corresponding drawing templet (either "rolling wheel" or a "sliding bar" display format). Boxes 352 to 354 then keeps track of the number of segments to draw from the CEA Scaling Array (S(J)). This array ranges from array element number 1 (S(1)) to array element number "Kend" (S(Kend)) where "Kend" is calculated via box 335 in FIG. 8. Box 355 determines the contents of each array element for the CEA Scaling Array (S(J)). If the value of a CEA Scaling Array element is equal to 1 (S(J)=1) then the corresponding segment in the drawing templet is set to 1 via box 357 (which specifies that a solid arc segment for a "rolling wheel" display format or a solid rectangular segment for a "sliding bar" display format is to be drawn). If the value of a CEA Scaling Array element is not equal to 1 (i.e. S(J)=0) then the corresponding segment in the drawing templet is set to 0 via box 356 (which specifies that a null arc segment for a "rolling wheel" display format or a null rectangular segment for a "sliding bar" display format is to be drawn). The selected drawing segment is then drawn on the CRT (item 66 on FIG. 1) via box 358. Box 359 reinitializes counting variable "J" to zero (0) after all "Kend" segments have been drawn as determined by box 354. The graphical displays provide the user with an easily understood representation of the accumulated time and accumulated EFPD exposure for CEA rod groups relative to the LCO's. The display formats are designed to present the data in terms of a contiguous monitoring interval using a spatial representation. FIG. 21 illustrates the format of the "Rolling Wheel" display. In this embodiment, two Part Strength CEA groups are assumed and a LCO limitation of no more than 5 EFPD exposure per 30 EFPD interval is specified (where exposure is defined as a Part Strength group being inserted between the Long Term Steady State Insertion Limit and the Transient Insertion Limit). The 30 EFPD interval is defined to be a contiguous 30 EFPD period. The contiguous 30 EFPD interval is depicted by rotating wheels; one for each Part Strength group. Each wheel rotates in a counterclockwise direction. A full rotation of a wheel (360 degrees) corresponds to the 30 EFPD contiguous monitoring interval. Shaded pie segments within a wheel represent the EFPD exposure for the Part Strength group whenever it was inserted between the Long Term Steady State Insertion Limit and the Transient Insertion Limit. As EFPD is accumulated, old exposure data is continuously discarded (rolls off the "Rolling Wheel"), while new data is continuously added (rolls on to the "Rolling Wheel"). Thus, the exposure of the Part Strength rod groups is maintained for a contiguous monitoring interval (window) using a spatial representation. FIG. 22 illustrates the format of the "Sliding Bar" display. As with FIG. 21, two Part Strength CEA groups are assumed and a LCO limitation of no more than 5 EFPD exposure per 30 EFPD interval is specified (where exposure is defined as a Part Strength group being inserted between the Long Term Steady State Insertion Limit and the Transient Insertion Limit). The 30 EFPD interval is defined to be a contiguous 30 EFPD period. The contiguous 30 EFPD interval is depicted by a linear line. The length of the line represents the contiguous 30 EFPD interval. There are two such linear lines, one for each Part Strength group. "Bars" , which are located above each line, represent the EFPD exposure for the Part Strength group whenever it was inserted between the Long Term Steady State Insertion Limit and the Transient Insertion Limit. The "Bars" slide along the line, moving from right to left. A full translation of the line by a "Bar" corresponds to a "Bar" fully transitioning the 30 EFPD contiguous monitoring interval. As EFPD is accumulated, old exposure data is continuously discarded ("Bars" or portions thereof slide off the line), while new data is continuously added ("Bars" or portions thereof slide on to the line). Thus, the exposure of the Part Strength rod groups is maintained for a contiguous monitoring interval (window) using a spatial representation. The Sector feature which is associated with the graphical displays (FIGS. 21 and 22) allows users to define "sectors" within the "Rolling Wheel" and "Sliding Bar" displays to be expanded and thus examined at higher resolutions. After the user enters the desired sector region to be expanded, the scales on the "Rolling Wheels" or "Sliding Bars" are rescaled to the range as entered by the user and the accumulate exposure data is displayed with proportionally greater resolution. The Query Mode (selected per FIGS. 21 and 22) allows the user to: (1) recall historic information and (2) to determine when a certain level of accumulated exposure (in terms of hours and/or EFPD) will "roll off/slide off" and be regained as usable margin (by having the user enter the future planned "power-time" profile for the plant). The Summary Display Mode (FIG. 23) provides an overall assessment of the current accumulated exposure and remaining margin for all Regulating and Part Strength rod groups utilizing a single convenient display page. Alarming capability (where is this depicted?) is provided to alert the user of an approach to an alarm condition (LCO), so that action may be taken prior to actually exceeding the alarm setpoint. In the event that the alarm setpoint is exceeded, an alarm annunciation alerts the user and provides a display (countdown clock) of the remaining time to take the prescribed corrective action relative to the required Completion Time as specified within the Technical Specifications for operation. In cases for which several corrective actions with differing Completion Time lines are specified, a count down clock representation for each corrective action is displayed. FIG. 11 depicts the functional logic for the HISTORIC DATA PLAYBACK MODULE. This module is activated whenever an operator selects the "Historic Data Option" function key on the keyboard (item 64 on FIG. 1) that is associated with the digital computer (item 54 on FIG. 1). The HISTORIC DATA PLAYBACK MODULE recalls previously archived CEA exposure historical data for playback and allows the operator to review CEA exposures from previous time intervals. The playback is for one day periods (as determined by the time stamp which is associated with each saved CEA exposure value). Box 300 prompts the operator to enter the "Start Time" point for the historic data and places the requested year and day for the playback into variables "YEAR" and "DAY". Box 301 determines if the "Start Time" requested by the operator is valid and within range of the existing historic data records (Limit D2 is either 364 days or 365 days dependent if "YEAR" corresponds to a "leap year" or not, or it is the current day if "YEAR" is equal to the current year. Limit D1 is either 1 if "YEAR" is greater than the first year of recorded archived records or it is equal to the first day in which the archived record exits if "YEAR" is equal to the earliest year of recorded archived records. Limit Y1 corresponds to the earliest year of recorded archived records and limit Y2 corresponds to the current year.). If an invalid time request is entered, box 302 rejects the request and displays an error message to the operator on the CRT (item 66 on FIG. 1). If the operator request is for a valid "Start Time", box 303 then recalls the historic archived CEA exposure record file (via the digital computer disk, item 62 on FIG. 1) based on the "YEAR" and "DAY" values. The program will use the first array element it encounters that begins on the requested day. Array elements are time stamped with the hour, day and year that they were recorded. The selected historic data is then formatted in a tabular format via box 304 and is output on the CRT (item 66 on FIG. 1). The playback of historic data is terminated when the operator selects a "Return to Real Time Data" option. This option is only displayed on the CRT when a Historic Data Playback is active. The "Return to Real Time Data" option is activated via a function key on the computer keyboard (item 64 on FIG. 1). The Predictor Mode (selected per FIGS. 21 and 22) allows the effect of a planned CEA Rod Group maneuver (for accumulated hours and/or EFPD exposure) to be assessed in advance of performing the actual maneuver. The user enters the planned Rod Group maneuver ("position-time" profile for the rod groups) and the anticipated plant "power-time" profile. The system then determines if there is sufficient margin (hours or EFPD exposure) to perform the maneuver based on the current exposure data and the information as entered by the user. If insufficient time (hours) or EFPD margin is available, the Predictor Mode projects when suitable margin will be regained to allow the maneuver to occur while maintaining compliance with the LCO's. An overview of the logic for the PREDICTOR MODULE is provided in FIG. 12. This module is called by the PROGRAM EXECUTIVE whenever a request is made for the Predictor Mode of operation. Requests are made via a function key on the digital computer keyboard, item 64 on FIG. 1. The PREDICTOR MODULE predicts if sufficient CEA exposure margin currently exists to perform a planned CEA maneuver. If insufficient margin exits, the PREDICTOR MODULE predicts when sufficient CEA exposure margin will be regained to perform the planned CEA maneuver. Referring to FIG. 12, an estimate of the CEA exposure for the planned CEA maneuver is entered by the operator, a determination is then made if sufficient CEA exposure margin currently exists to perform the planned CEA maneuver. If sufficient CEA exposure margin currently exists, a message is output to the operator indicating that sufficient margin exists to perform the planned maneuver. If insufficient CEA exposure margin currently exists, then the program predicts when sufficient margin will be regained to perform the planned maneuver. FIG. 13 depicts the basic logic for the PREDICTOR MODULE. In box 400, the value of the estimated value of CEA exposure for the planned maneuver is entered by the operator and stored in variable "Delta_Margin". Box 401 next obtains the current value of CEA exposure ("EXPOSURE") as last computed by the CEA EXPOSURE CALCULATION MODULE (box 305 of FIG. 7) which is stored in the data base. Box 402 then determines the total required CEA exposure which would occur if the planned CEA maneuver is performed at the current point in time ("Required_Margin" ). This total required CEA exposure margin is the sum of the current CEA exposure ("EXPOSURE") and the estimated CEA exposure to perform the planned maneuver ("Delta_Margin"). Next, box 403 determines if there is sufficient margin to perform the planned CEA maneuver by comparing variable "Required_Margin" to the LCO ("LIMIT"). The LCO limit is stored in the data base of the digital computer (item 60 of FIG. 1). If there is presently sufficient total CEA exposure margin to perform the planned maneuver then box 405 outputs a message to the operator on the CRT indicating that sufficient margin exists to perform the planned maneuver. If insufficient total CEA exposure margin currently exists, then box 404 predicts when sufficient margin will be regained to perform the planned maneuver by executing the MARGIN PREDICTION MODULE which is illustrated in FIG. 14. With reference to FIG. 14 (MARGIN PREDICTION MODULE), box 420 first maps the Accumulated CEA Exposure data, stored in the CEA Exposure Data Array (array "P(J)") into a second array (array "PP(J)") which is then further utilized in the MARGIN PREDICTION MODULE to predict when CEA exposure margin will be regained. For simplicity, the logic is illustrated for the case of a single program array--however, all requisite data arrays associated with the CEA Rod Position System are similarly processed. Array "PP(J)" is used to avoid altering data in array "P(J)" while performing the prediction computations. The logic for the Array Mapping is now explained. After this logic is described, an explanation of the MARGIN PREDICTION MODULE, FIG. 14, will resume. The Array Mapping is illustrated in FIG. 15 (ARRAY MAPPING MODULE). Box 450 initiates the logic by setting the counting variable "J" to zero. Boxes 451 and 452 determine when all the array elements from array "P(J)" have been mapped into array "PP(J)". The completion of the array mapping occurs when counting variable "J" is greater than the highest numbered array element in the CEA Exposure Data Array (array element number "N"). Box 453 performs the mapping by setting the value of array element "PP(J)" to the value of array element "P(J)". When all the array elements from array "P(J)" have been mapped into array "PP(J)" then box 454 resets counting variable "J" to zero. Returning to the MARGIN PREDICTION MODULE of FIG. 14, after the array mapping, then box 421 determines the maximum allowable accumulated CEA exposure which would still allow the planned CEA maneuver to occur by computing the difference between the LCO margin limit ("LIMIT") and the CEA exposure required to perform the planned CEA maneuver ("Delta--Margin"). This value is stored in variable ("Max_Exposure") and represents the amount of accumulated CEA exposure that can exist prior to beginning the planned CEA maneuver (values of CEA exposure which are larger than this amount will result in insufficient CEA exposure margin to perform the planned maneuver; i.e.--the sum of the CEA exposure required to perform the maneuver and the current value of CEA exposure is such that they collectively exceed the LCO as specified in variable "LIMIT"). Boxes 422 to 426 then continually update array "PP(J)" until sufficient CEA exposure margin is lost (as determined via variable "New_Exposure"). This is determined as follows: a value of zero is inserting into position "PP(1)" (which represents the most recent 1 EFP-hour interval); each element of array "PP(J)" is then upward shifted by one position; and finally the value of the last array element "PP(N)" is deleted. This process simulates plant operation with all CEA rods above the Steady State and Transient Insertion Limits (for this condition, no CEA exposure accumulation occurs). Box 422 initializes the counting variable to zero while box 423 accumulates the number of simulated 1 EFP-hour intervals with zero CEA exposure accumulation (this is equivalent to the number of program passes for the "Margin Prediction Calculation"). For each program pass, box 424 (UPDATE CEA EXPOSURE FOR ARRAY PP MODULE) updates the CEA exposure for array "PP" (zero CEA exposure for the most recent simulated 1 EFP-hour interval) and box 425 (NEW_MARGIN CALCULATION MODULE) calculates the corresponding new value of accumulated CEA exposure (which is stored in variable "New_Exposure"). Program passes are continually made (each pass representing 1 EFP-hour interval with no CEA exposure accumulation) until sufficient old margin "rolls off" and the remaining accumulated CEA exposure ("New_Exposure") is sufficiently reduced to allow the planned maneuver, as determined by box 426. The logic for the UPDATE CEA EXPOSURE FOR ARRAY PP MODULE is now explained. After this logic is described, explanation of the MARGIN PREDICTION MODULE, FIG. 14, will resume. The UPDATE CEA EXPOSURE FOR ARRAY PP MODULE is illustrated in FIG. 16. When called, this module updates (shifts up by one position) each element of array "PP(J)" and inserts a value of zero into the first array position ("PP(1)"). Thus, the first position of the array "PP(1)" is updated with a value of zero, the second array position is updated with the value from the first array position, etc. until all CEA exposure data has been shifted up 1 EFP-hour interval in the array. The last array element "N" (which represents the oldest CEA exposure data) is discarded since it is now beyond the LCO EFP duration criteria. This shifting of elements of array "PP(J)" represents a 1 EFP-hour interval of operation with zero CEA exposure accumulation. Box 460 first initializes the counting variable. Boxes 461 to 463 shift up the data array elements, beginning from the last array position (that is, first array element "N-1" is shifted into array position "N", then array element "N-2" is shifted into array position "N-1", etc.) until array element 1 (the last shift performed by boxes 461 to 463 is from array position 1 to array position 2). After array position 1 is shifted into position 2, box 464 inserts a value of zero (0) into array position "PP(1)". This shifting which occurs during a single program pass represents 1 EFP-hour interval of operation with zero (0) CEA exposure accumulation. Returning now to the MARGIN PREDICTION MODULE of FIG. 14, after execution of the UPDATE CEA EXPOSURE FOR ARRAY PP MODULE is completed, then box 425 determines the new value of CEA exposure for array "PP(J)" which is calculated by the NEW_MARGIN CALCULATION MODULE. The logic for the NEW--MARGIN CALCULATION MODULE is now explained. After this logic is described, explanation of the MARGIN PREDICTION MODULE, FIG. 14, will resume. The NEW_MARGIN CALCULATION MODULE is illustrated in FIG. 17. This module determines the value of CEA exposure for array "PP(J)". This module is called by the MARGIN PREDICTION MODULE immediately after the elements of array "PP(J)" have been shifted up by one position (equivalent to a 1 EFP-hour interval with no CEA exposure accumulation). Referring to FIG. 17, boxes 480 to 485 calculate the current value of CEA exposure in array "PP(J)" by summing the contents of each array element. The contents of each array element contains the value of CEA exposure for a given 1 EFP-hour interval. The total accumulated CEA exposure is then stored in variable "SUM" via box 484 after all "N" array elements are added (as determined via box 483). The module then sets variable "New_Exposure" to variable "SUM" in box 485. Returning to the MARGIN PREDICTION MODULE of FIG. 14, after the value of "New--Exposure" is determined, box 426 then determines if the value of "New--Exposure" is greater than the value of "Max--Exposure". If true, then insufficient CEA exposure margin has yet to "roll off" (the present CEA exposure accumulation as stored in variable "New_Exposure" is such that there is insufficient available margin to accommodate the "Delta_Margin" and remain within the LCO as defined in variable "LIMIT") and the module then begins another program pass (equivalent to another 1 EFP-hour interval with zero CEA exposure) by returning to box 423. When sufficient CEA exposure margin has "rolled off" (false condition for box 426) then the total number of EFP intervals to achieve the reduction (as contained in counting variable "I") is then stored in variable "INTERVALS" via box 427. Variable "INTERVALS" therefore represents the number of hours in which sufficient CEA exposure margin will be regained to perform the planned maneuver, assuming the plant operates at a power rating of 100% EFP. Boxes 428 to 430 convert "INTERVALS" into equivalent time in terms of "days" and "hours" and box 431 translates this time interval into calendar time. Box 432 displays the predicted time (at a power condition of 100% EFP) when sufficient CEA exposure margin will be regained to perform the planned CEA maneuver. An example of the displayed output would be as follows: "Sufficient Margin will exist after "DD" days and "HH" hours of operating at a power level of 100% EFP which corresponds to "DAY, MONTH, YEAR and TIME". If operating at less than 100% EFP, then the time period will be proportional to the value of EFP relative to 100% EFP e.g., "(100% EFP)/(actual EFP)". Finally, box 433 resets the counting variable to zero. In the event of a system outage, the Update Mode allows the system to be recalibrated to the current operational conditions. After the system is brought on-line, the user enters the appropriate plant "time-power" profile and the rod group exposure profile for the outage interval. Based on this information and the rod exposure information stored up to the time of the outage, the system is recalibrated to the current operational conditions and restored to operational service. FIG. 18 depicts the basic logic for the UPDATE MODULE. This module is called by the PROGRAM EXECUTIVE whenever there is a computer restart or whenever requested by the operator via a function key on the keyboard (item 64 on FIG. 1) associated with the digital computer. Box 500 first requests that the operator confirm that an Update is to be performed. If the operator enters a "yes" to this prompt (using the keyboard) then, via box 501, the operator is prompted to enter the CEA exposure history for the outage period. The operator enters this information using the keyboard in conjunction with an update data templet which appears on the CRT screen (item 66 in FIG. 1). For each hourly EFP interval which occurred during the outage, the operator determines if there was any CEA exposure for that interval. For hourly EFP intervals in which there was no CEA exposure, the operator enters zero. For hourly EFP intervals in which there was CEA exposure, the operator enters the value of the CEA exposure which occurred. If the rods were inserted for the full time during a given hourly EFP interval then the CEA exposure corresponds to 1 EFP-hour; if they were inserted for only a portion of the hourly EFP interval, then the CEA exposure would correspond to a fraction of 1 EFP-hour. The operator enters this information for each affected CEA rod group. If a CEA rod group was not inserted during the outage interval, the operator enters a "not inserted" command and the computer sets the CEA exposure for all the hourly EFP intervals to zero for that CEA rod group. This allows a quick update for such cases so the operator need not manually insert a zero for each 1 EFP-hour interval. The operator estimates the EFP information based on written logs of plant power and rod positions for the outage period. Upon completion of the operator entry of the CEA exposure history for the outage period, variable "KNUMBER" stores the value of the total number of hourly EFP intervals which occurred during the outage period and array "E(J)" stores the value of CEA exposure (as entered by the operator) for each of the hourly EFP intervals which occurred during the outage. Boxes 502 to 505 then shift the elements of array "P(J)" up by one position and inserts a value of "E(m)" for into the first array element, thus updating the CEA exposure history for a 1 EFP-hour interval. Box 504 determines when the shift has been completed (when condition "m>KNUMBER" is true). The logic for box 505 is now explained. After this logic is described, explanation of the UPDATE MODULE, FIG. 18, will resume. The HOURLY EFP INTERVAL UPDATE MODULE is illustrated in FIG. 19. When called, this module updates (shifts up by one position) each element of array "P(J)" and inserts the CEA exposure value as entered into "E(m)" into the first array position ("P(1)"). Thus, the first position of the array "P (1)" is updated with the value of CEA exposure as was entered by the operator into "E(m)" for this EFP hourly interval, the second array position is updated with the value from the first array position, etc. until all CEA exposure data has been shifted up 1 EFP-hourly interval in the array. The last array element "N" (which represents the oldest CEA exposure data) is discarded. This shifting of elements of array "P(J)" up by one position represents a 1 EFP hourly update interval. Referring again to FIG. 19, box 520 first initializes the counting variable. Boxes 521 to 523 shift up the data array elements, beginning from the last array position (that is, first array element "N-1" is shifted into array position "N", then array element "N-2" is shifted into,array position "N-1", etc.) until array element 1 (the last shift performed by boxes 521 to 523 is from array position 1 to array position 2). After array position 1 is shifted into position 2, box 524 inserts the value of CEA exposure as was entered by the operator into "E(m)" for this EFP hourly interval. Thus array "P(J)" is updated for a 1 EFP hourly period. With further reference to the UPDATE MODULE of FIG. 18; after the elements of array "P(J)" have been updated for 1 EFP hourly interval, then boxes 503 and 504 determine when the array has been updated for all of the EFP hourly periods which occurred during the outage. This occurs when box 504 determines that the condition "m>KNUMBER" is true. Box 506 then resets the counting variable to zero. From the foregoing, it can be appreciated that the invention has been described in the context of a nuclear power plant having a nuclear reactor core and a multiplicity of control rods arranged for movement through the core for controlling the reactor power. This multiplicity of rods includes a plurality of groups of control rods, the groups being movable through the core in staggered sequence. Each group is subject to an administrative limit on the cumulative exposure in the core while each group is situated within a pre-established position range in the core. Independent of any symbology utilized in connection with FIGS. 1-20 and associated description hereinabove, one administrative limit can be expressed in the form of a limit index W:X defined by a maximum of W hours of accumulated effective exposure on the sum S of effective exposure occurrences W.sub.1, W.sub.2 . . . during any X hour reference period, with W<X. Increments in the associated time base, are one hour each. Another form of the administrative limit can be expressed as a limit index Y:Z defined by a maximum of Y effective full power hours of exposure consisting of the sum of effective exposure occurrences Y.sub.1,Y.sub.2 . . . during any Z hour reference period of effective full power operation of the core, with Y<Z. Increments in the associated time base, are one effective full power hour each. Yet another administrative limit can be the maximum permitted time interval T during which a group can be positioned between, e.g., the Short Term Steady State Insertion Limit and the maximum insertion position which is permitted during a normal operational transit. The invention also includes a novel form of displaying the comparison of the accumulated effective exposure for each group with the administrative limit for each group as shown in FIGS. 21-23. In terms of the symbology described immediately above, the display 600,600' of FIGS. 21 and 22 includes at least one scale 601,601' of Z uniform intervals 602,602', marked by a plurality of numeric values 603,603' indicative of an initial zero value 604,604' and a final value Z 605,605'. An indicator configuration 606,606' for each group is displayed, each indicator configuration having a scale associated therewith, and consisting of an indicator 607,607' for each component y of the sum S. Each indicator initially appears at the zero representation of the scale and grows in size toward the scale value Z to span the number of scale intervals corresponding to the ratio of effective exposure of component y to the effective power interval Z. Independently of but simultaneously with the indicator growth, each indicator along the scale advances toward the scale value Z, at a uniform rate. The sum S of all components y during the immediately preceding interval is displayed 608,608' adjacent to the scale Z. Instantaneous margin M=Y-S, can also be displayed 609,609'. Thus at any given moment, the operator can visually recognize the number of and effective exposure for each component y during the immediately preceding core effective full power interval Z; the total exposure of S during the immediately preceding interval of Z; and the margin M. In one embodiment, as shown in FIG. 21, a respective scale 601 is displayed for each group. Each scale is displayed as a circle with coincident zero and Z values. Each indicator 607 of a component y is displayed as a sector of the circle, which grows by increasing the included angle of the sector and which advances by continually rotating about the center of the circle toward the value Z. Another display embodiment is shown in FIG. 22. One scale 601' is displayed as a linear segment with the zero value 604' at one end and the Z value 605' at the other end. The indicator configuration 606',606" for each of at least two groups is associated with the one scale. Each indicator 607' of a component y is displayed as a horizontal bar, which grows by increasing in horizontal length, and which advances by continually moving horizontally toward the value Z. Similar displays can be presented for monitoring a limit expressed by W:X. FIG. 23 shows a summary report in a tabular form.
	description	This application is related to U.S. patent application Ser. No. 12/986,217, entitled SELF-POWERED WIRELESS IN-CORE DETECTOR, filed Jan. 7, 2011, concurrently herewith, now U.S. Pat. No. 8,681,920. 1. Field of the Invention The present invention pertains generally to apparatus for monitoring the radiation within the core of a nuclear reactor and, more particularly, to such apparatus that will not obstruct refueling of the reactor. 2. Related Art In many state-of-the-art nuclear reactor systems in-core sensors are employed for measuring the radioactivity within the core at a number of axial elevations. These sensors are used to measure the radial and axial distribution of the power inside the reactor core. This power distribution measurement information is used to determine whether the reactor is operating within nuclear power distribution limits. The typical in-core sensor used to perform this function is a self-powered detector that produces an electric current that is proportional to the amount of fission occurring around it. This type of sensor does not require an outside source of electrical power to produce the current and is commonly referred to as a self-powered detector and is more fully described in U.S. Pat. No. 5,745,538, issued Apr. 20, 1998, and assigned to the Assignee of this invention. FIG. 1 provides a diagram of the mechanisms that produce the current I(t) in a self-powered detector element 10. A neutron sensitive material such as vanadium is employed for the emitter element 12 and emits electrons in response to neutron irradiation. Typically, the self-powered detectors are grouped within instrumentation thimble assemblies. A representative in-core instrumentation thimble assembly is shown in FIG. 2. The signal level generated by the essentially non-depleting neutron sensitive emitter element 12 shown in FIG. 1, is low, however, a single, full core length neutron sensitive emitter element provides an adequate signal without complex and expensive signal processors. The proportions of the full length signal generated by the single neutron sensitive emitter element attributable to various axial regions of the core are determined from apportioning the signal generated by different lengths of gamma sensitive elements 14 which define the axial regions of the core and are shown in FIG. 2. The apportioning signals are ratioed which eliminates much of the effects of the delayed gamma radiation due to fission products. The in-core instrumentation thimble assemblies also include a thermocouple 18 for measuring the temperature of the coolant exiting the fuel assemblies. The electrical signal output from the self-powered detector elements and the thermocouple in each in-core instrumentation thimble assembly in the reactor core are collected at the electrical connector 20 and sent to a location well away from the reactor for final processing and use in producing the measured core power distribution. FIG. 3 shows an example of a core monitoring system presently offered for sale by Westinghouse Electric Company LLC with the product name WINCISE™ that employs fixed in-core instrumentation thimble assemblies 16 within the instrument thimbles of fuel assemblies within the core to measure the core's power distribution. Cabling 22 extends from the instrument thimble assemblies 16 through the containment seal table 24 to a signal processing cabinet 26 where the outputs are conditioned, digitized and multiplexed and transmitted through the containment walls 28 to a computer workstation 30 where they can be further processed and displayed. The thermocouple signals from the in-core instrumentation thimble assemblies are also sent to a reference junction unit 32 which transmits the signals to an inadequate core cooling monitor 34 which communicates with the plant computer 36 which is also connected to the workstation 30. Because of the hostile environment, the signal processing cabinet 26 has to be located a significant distance away from the core and the signal has to be sent from the detector 16 to the signal processing cabinet 26 through specially constructed cables that are extremely expensive and the long runs reduce the signal to noise ratio. Unfortunately, these long runs of cable have proved necessary because the electronics for signal processing has to be shielded from the highly radioactive environment surrounding the core region. In previous nuclear plant designs, the in-core detectors entered the reactor vessel from the lower hemispherical end and entered the fuel assemblies instrumentation thimble from the bottom fuel assembly nozzle. In at least some of the current generation of nuclear plant designs, such as the AP1000 nuclear plant, the in-core monitoring access is located at the top of the reactor vessel, which means that during refueling all in-core monitoring cabling will need to be removed before accessing the fuel. A wireless in-core monitor that is self-contained within the fuel assemblies and wirelessly transmits the monitored signals to a location remote from the reactor vessel would allow immediate access to the fuel without the time-consuming and expensive process of disconnecting, withdrawing and storing the in-core monitoring cables before the fuel assemblies could be accessed, and restoring those connections after the refueling process is complete. A wireless alternative would thus save days in the critical path of a refueling outage. A wireless system also allows every fuel assembly to be monitored, which significantly increases the amount of core power distribution information that is available. However, a wireless system requires that electronic components be located at or very near the reactor core where gamma and neutron radiation and high temperatures would render semiconductor electronics inoperable within a very short time. Vacuum tubes are known to be radiation insensitive, but their size and current demands have made their use impractical until recently. Recent developments in micro-electromechanical devices have allowed vacuum tubes to shrink to microscopic sizes and significantly reduced power draw demands. Accordingly, it is an object of this invention to improve the critical path for refueling a reactor by significantly reducing the number of cables attached to the reactor head that would have to be removed and reconnected in the course of the refueling process. It is a further object of this invention to provide a fuel assembly with a self-contained instrument thimble assembly that can be inserted into the core of a nuclear reactor and placed in operation without the necessity of routing cabling and connectors through the reactor vessel to activate the instrumentation. It is an additional object of this invention to increase the amount of in-core power distribution data that is communicated to the plant operator. These and other objectives are achieved by the apparatus of this invention which avoids the necessity of running expensive electrical cables through the reactor head and reactor internals to connect with and energize the in-core instrumentation. In accordance with this invention, a nuclear reactor in-core detector system is provided, including an in-core nuclear instrumentation thimble assembly that is substantially wholly contained within an instrument thimble within a nuclear fuel assembly. The instrument thimble assembly includes a self-powered, fixed, in-core detector for monitoring a reactor core parameter indicative of a state of the reactor core and providing an electric output representative of the monitored parameter. The instrument thimble assembly also includes a wireless transmitter that is connected to receive the electrical output from the self-powered fixed in-core detector and wirelessly transmit that signal to a location outside the reactor. Desirably, the wireless transmitter comprises a number of electronic components at least one of which is a vacuum microelectronic device and, preferably, a vacuum diode placed in a grid circuit of an amplifier which is connected to the electrical output of the self-powered, fixed, in-core detector and responds substantially logarithmically, thus enabling the electronic components to follow the monitored neutron flux from start-up to full power of a nuclear reactor in which the in-core detector system is disposed. In another embodiment, in addition to the amplifier, the electronics components include a current-to-voltage converter and a voltage controlled oscillator with an output of the amplifier connected to an input of the current-to-voltage converter whose output is connected to an input of the voltage controlled oscillator that provides a frequency output proportional to a voltage on the input of the voltage controlled oscillator. In that way, the current which is the electrical output representative of the monitored parameter, which is connected to the amplifier, is converted to a corresponding frequency signal that can be transmitted by a wireless transmitter. In still another embodiment, the voltage controlled oscillator comprises a micro-electronic reactance tube. Preferably, the electronic components comprise—an input of a first amplifier connected to the electrical output of the self-powered, fixed, in-core detector; the input of the current-to-voltage converter connected to an output of the amplifier; the input of the voltage controlled oscillator connected to the output of the current-to-voltage converter; an input of a second amplifier connected to the output of the voltage controlled oscillator; and a wireless transmission circuit connected to an output of the second amplifier for wirelessly transmitting the output of the second amplifier. Desirably, the nuclear reactor in-core detector system includes a wireless receiver circuit and signal conditioning component designed to be situated outside the highly radioactive environment of the nuclear reactor containment, and preferably, including conventional solid state components. In still another embodiment, the nuclear reactor in-core detector system includes a wireless receiver positioned outside and within the vicinity of the reactor vessel for receiving signals from the wireless transmitter and a retransmitter for transmitting outside the containment the signals received from the wireless transmitter. Desirably, the retransmitter is a second wireless transmission circuit that transmits the signals received from the wireless transmitter to a second wireless receiver that communicates the signals received from the wireless transmitter, by way of the wireless receiver and the retransmitter, to processing circuitry outside the containment. Desirably, the second wireless receiver is positioned within the vicinity of a containment wall that shields the primary circuit of a nuclear power generation facility in which the in-core detector system is placed. In a further embodiment, the invention comprises a nuclear fuel assembly having a top nozzle and a bottom nozzle and a plurality of thimble tubes extending between and substantially connected to the top nozzle and the bottom nozzle. At least one of the thimble tubes comprises an instrumentation thimble that houses and substantially completely contains the fixed in-core monitoring component of the detector system of this invention. The primary side of nuclear power generating systems which are cooled with water under pressure comprises a closed circuit which is isolated from and in heat exchange relationship with a secondary side for the production of useful energy. The primary side comprises the reactor vessel enclosing a core internal structure that supports a plurality of fuel assemblies containing fissile material, the primary circuit within heat exchange steam generators, the inner volume a pressurizer, pumps and pipes for circulating pressurized water, the pipes connecting each of the steam generators and pumps to the reactor vessel independently. Each of the parts of the primary side comprising a steam generator, a pump and a system of pipes which are connected to the vessel form a loop of the primary side. For the purpose of illustration, FIG. 4 shows a simplified nuclear reactor primary system, including a generally cylindrical reactor pressure vessel 40 having a closure head 42 enclosing a nuclear core 44. A liquid reactor coolant, such as water, is pumped into the vessel 40 by pump 46 through the core 44 where heat energy is absorbed and is discharged to a heat exchanger 48, typically referred to as a steam generator, in which heat is transferred to a utilization circuit (not shown), such a steam driven turbine generator. The reactor coolant is then returned to the pump 46 completing the primary loop. Typically, a plurality of the above-described loops are connected to a single reactor vessel 40 by reactor coolant piping 50. An exemplary reactor design incorporating this invention is shown in FIG. 5. In addition to the core 44 comprised of a plurality of parallel, vertical, co-extending fuel assemblies 80, for purposes of this description, the other vessel internal structures can be divided into the lower internals 52 and the upper internals 54. In conventional designs, the lower internals' function is to support, align and guide core components and instrumentation as well as direct flow within the vessel. The upper internals 54 restrain or provide a secondary restraint for the fuel assemblies 80 (only two of which are shown for simplicity in this figure), and support guide instrumentation and components, such as control rods 56. In the exemplary reactor shown in FIG. 5, coolant enters the reactor vessel 40 through one or more inlet nozzles, flows down through an annulus between the vessel 40 and the core barrel 60, is turned 180° in a lower reactor vessel plenum 61, passes upwardly through a lower support plate and a lower core plate 64, upon which the fuel assemblies 80 are seated, and through and about the assemblies. In some designs, the lower support plate 62 and the lower core plate 64 are replaced by a single structure, the lower core support plate that has the same elevation as 62. Coolant exiting the core 44 flows along the underside of the upper core plate 66 and upwardly and through a plurality of perforations 68 in the upper core plate 66. The coolant then flows upwardly and radially to one or more outlet nozzles 70. The upper internals 54 can be supported from the vessel 40 or the vessel head 42 and includes an upper support assembly 72. Loads are transmitted between the upper support assembly 72 and the upper core plate 66 primarily by a plurality of support columns 74. Each support column is aligned above a selected fuel assembly 80 and perforation 68 in the upper core plate 66. The rectilinearly movable control rods 56 typically include a drive shaft 76 and a spider assembly 78 of neutron poison rods that are guided through the upper internals 54 and into aligned fuel assemblies 80 by control rod guide tubes 79. FIG. 6 is an elevational view represented in vertically shortened form, of a fuel assembly being generally designated by reference character 80. The fuel assembly 80 is the type used in a pressurized water reactor and has a structural skeleton which at its lower end includes a bottom nozzle 82. The bottom nozzle 82 supports the fuel assembly 80 on the lower core support plate 64 in the core region of the nuclear reactor. In addition to the bottom nozzle 82, the structural skeleton of the fuel assembly 80 also includes a top nozzle 84 at its upper end and a number of guide tubes or thimbles 86, which extend longitudinally between the bottom and top nozzles 82 and 84 and at opposite ends are rigidly attached thereto. The fuel assembly 80 further includes a plurality of transverse grids 88 axially spaced along and mounted to the guide thimbles 86 (also referred to as guide tubes) and an organized array of elongated fuel rods 90 transversely spaced and supported by the grids 88. Although it cannot be seen in FIG. 6, the grids 88 are conventionally formed from orthogonal straps that are interleaved in an egg-crate pattern with the adjacent interface of four straps defining approximately square support cells through which the fuel rods 90 are supported in transversely spaced relationship with each other. In many conventional designs, springs and dimples are stamped into the opposing walls of the straps that form the support cells. The springs and dimples extend radially into the support cells and capture the fuel rods therebetween; exerting pressure on the fuel rod cladding to hold the rods in position. Also, the assembly 80 has an instrumentation tube 92 located in the center thereof that extends between and is mounted to the bottom and top nozzles 82 and 84. With such an arrangement of parts, the fuel assembly 80 forms an integral unit capable of being conveniently handled without damaging the assembly of parts. As mentioned above, the fuel rods 90 in the array thereof in the assembly 80 are held in spaced relationship with one another by the grids 88 spaced along the fuel assembly length. Each fuel rod 90 includes a plurality of nuclear fuel pellets 94 and is closed at its opposite ends by upper and lower end plugs 96 and 98. The fuel pellets 94 are maintained in a stack by a plenum spring 100 disposed between the upper end plug 96 and the top of the pellet stack. The fuel pellets 94, composed of fissile material, are responsible for creating the reactive power of the reactor. The cladding, which surrounds the pellets, functions as a barrier to prevent the fission byproducts from entering the coolant and further contaminating the reactor systems. To control the fission process, a number of control rods 56 are reciprocably movable in the guide thimbles 86 located at predetermined positions in the fuel assembly 80. Specifically, a rod cluster control mechanism (also referred to as the spider assembly) 78 positioned above the top nozzle 84 supports the control rods 56. The control mechanism has an internally threaded cylindrical hub member 102 with a plurality of radially extending flukes or arms 104 that with the control rods 56 form the spider assembly 78 that was previously mentioned with respect to FIG. 5. Each arm 104 is interconnected to the control rods 56 such that the control mechanism 78 is operable to move the control rods vertically in the guide thimbles 86 to thereby control the fission process in the fuel assembly 80, under the motor power of control rod drive shaft 76 (shown in FIG. 5) which are coupled to the control rod hubs 102, all in a well known manner. As mentioned above, in the AP1000 nuclear plant design, the in-core monitoring access is through the top of the reactor vessel, which is a significant departure from previous designs which fed the fixed in-core detector cables through the bottom of the vessel and into the fuel assembly instrument thimbles through the lower fuel assembly nozzle. The change in design means that during refueling all conventional in-core monitoring cabling will need to be removed before accessing the fuel. This invention provides a wireless in-core monitor that is wholly contained within the instrument thimble within the fuel assemblies without any tether that extends outside the core and would permit access to the fuel without going through the costly and time-consuming steps of removing and reconnecting the cabling. In accordance with this invention, the in-core instrument thimble assembly is illustrated as a block diagram in FIG. 7 and includes, in addition to the fixed in-core neutron detector, a self-contained power source and a wireless transmission circuit. Within the transmission circuit, the neutron detector output current is fed directly into an amplifier 112, thus eliminating cabling concerns. One or more stages of amplification are provided within the amplifier 112, using vacuum micro-electronic devices. A vacuum diode is preferably placed in the grid circuit of the amplifier to make the amplifier respond logarithmically, thus enabling the electronics to follow the neutron flux from start-up through full power. The amplified signal is then fed to a current-to-voltage converter 114. The output voltage of the current-to-voltage converter 114 is used as the input to a voltage controlled oscillator 118 which converts the voltage input to a frequency output. As the neutron flux changes, so will the voltage input to the voltage controlled oscillator, which will vary the output frequency. A vacuum micro-electronic reactance tube can be used for the voltage controlled oscillator 118. Such an arrangement provides a precise correlation between the neutron flux monitored by the neutron detector 10 and the output frequency of the voltage controlled oscillator 118. That output is then amplified by amplifier 120 whose output is communicated to a wireless transmitter 122 within the in-core instrument thimble assembly 16. The in-core instrument thimble assembly 16 can be made up of a single unit housing the neutron detector, power supply and transmission circuit or it can be made up of modular units, e.g., the self-contained power supply, neutron detector and transmission circuit, respectively. The primary electrical power source for the signal transmitting electrical hardware is the rechargeable battery 132 shown as part of the exemplary power supply illustrated in FIG. 8. The charge on the battery 132 is maintained by the use of the electrical power produced by a dedicated power supply self-power detector element 134 that is contained within the power supply 130, so that the nuclear radiation in the reactor is the ultimate power source for the device, keeping the battery 132 charged. The power supply self-powered detector element 134 is connected to the battery 132 through a conditioning circuit 136 and the battery is in turn connected to the signal transmitter circuit 138 that transmits the signal received from the fixed in-core detector and the thermocouple monitoring the core such as was described with respect to FIGS. 2A, 2B and 2C. The self-contained power supply is more fully described in U.S. patent application Ser. No. 12/986,217. FIG. 9 shows a schematic layout of a self-powered wireless in-core detector instrumentation core power distribution measurement system constructed in accordance with this invention. The schematic layout illustrated in FIG. 9 is identical to the schematic layout illustrated in FIG. 3 for a conventional in-core monitoring system, except that the in-core instrument thimble assembly has been rotated 180° so that the electrical connectors for the detector element are closer to a receiver of the wireless transmitted signal and the cabling has been replaced by the wireless transmitters and receivers 122, 124, 138 and 116, the in containment electronics 26 and 32 have been respectively replaced by the SPD signal processing system 108 and the core exit thermocouple signal processing system 106, located outside the containment 28. In all other respects, the systems are the same. As can also be appreciated from FIG. 9, the signal from the in-core instrument thimble assembly 16 wireless transmitter 122 is received by an antenna 124 on the underside of the reactor vessel head 42 which communicates with a combination wireless receiver and retransmitter 138 on the reactor head 42. In that way, the reactor head 42 can be removed and the fuel assemblies accessed without the in-core instrumentation being an obstacle. Placement of the transmitting antenna on the reactor vessel will depend on the reactor design but the intent is to transmit from a close proximity to the reactor vessel at a location that would not be an impediment to accessing the fuel assemblies. The neutron signal is then retransmitted by the retransmission circuit 138 to a receiver 116 proximate the containment outer wall. The combination receiver and retransmitter 138 should similarly be constructed from vacuum microelectronic devices because of their close proximity to the reactor vessel; however, the receivers 116 and the processing circuitry 106 and 108 can be constructed from conventional solid state components and may be located within the containment remote from the reactor vessel or outside the containment. Thus, this invention greatly simplifies the transmission of the in-core detector signals and the refueling operation. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
	abstract	The method of producing this diffraction grating includes a step of generating a moire by a periodic pattern projected onto a plurality of unit diffraction gratings and a plurality of unit diffraction gratings, and a step of adjusting so that the extending directions of the gratings are aligned by relatively rotating at least one of a plurality of unit diffractions with respect to at least one of the others of the plurality of unit diffractions.
	abstract	A pressurized water reactor (PWR) includes a vertical cylindrical pressure vessel having a lower portion containing a nuclear reactor core and a vessel head defining an integral pressurizer. A reactor coolant pump (RCP) mounted on the vessel head includes an impeller inside the pressure vessel, a pump motor outside the pressure vessel, and a vertical drive shaft connecting the motor and impeller. The drive shaft does not pass through the integral pressurizer. The drive shaft passes through a vessel penetration of the pressure vessel that is at least large enough for the impeller to pass through.
	claims	1. A positive/negative phase shift bimetallic zone plate; comprising:a first metallic material having a positive phase shift;a second metallic material having a negative phase shift at a working energy point;wherein the first metallic material and the second metallic material are alternately arranged, so that the second metallic material replaces the blank portion in a cycle of a traditional zone plate. 2. The positive/negative phase shift bimetallic zone plate of claim 1, wherein the positive/negative phase shift bimetallic zone plate is annular, and the first metallic material and the second metallic material form a structure of alternate rings. 3. The positive/negative phase shift bimetallic zone plate of claim 1, the first metallic material is selected from nickel, gold, germanium, titanium, vanadium, chromium, manganese, iron, copper, zinc. 4. The positive/negative phase shift bimetallic zone plate of claim 1, the second metallic material is selected from titanium, vanadium, chromium, manganese, iron, cobalt, nickel, copper, zinc, gallium, germanium, hafnium, tungsten, rhenium and osmium. 5. The positive/negative phase shift bimetallic zone plate of claim 1, wherein in the case that the positive/negative phase shift bimetallic zone plate has the same thickness as that of a normal monometallic phase zone plate, the diffraction efficiency of the positive/negative phase shift bimetallic zone plate is higher than the diffraction efficiency of the normal monometallic phase zone plate in conventional ranges. 6. The positive/negative phase shift bimetallic zone plate of claim 1, the positive/negative phase shift bimetallic zone plate is a vanadium-nickel, titanium-nickel, or vanadium-gold bimetallic zone plate. 7. A method of producing a positive/negative phase shift bimetallic zone plate, comprising following steps:a. depositing a thin film of a first metallic material on a substrate;b. forming a photoresist having a zone plate structure on the thin film of the first metallic material;c. transferring the zone plate structure to the thin film of the first metallic material by performing etching via the formed photoresist having the zone plate structure, so as to form a zone plate structure of the first metallic material;d. depositing the second metallic material at interspaces formed by the etching;e. removing the photoresist, so as to form a positive/negative phase shift bimetallic zone plate structure. 8. The method of claim 7, wherein the photoresist is coated by spin coating, and thereafter is subjected to electron beam exposureor interference lithography, so as to form a photoresist having a zone plate structure. 9. The method of claim 7, wherein the etching in step d is performed by argon ion etching or reactive ion etching. 10. The method of claim 7, further comprising:opening a window on the back side of the positive/negative phase shift bimetallic zone plate structure obtained in step e, to obtain the positive/negative phase shift bimetallic zone plate.
	summary
	abstract	Grids and collimators, for use with electromagnetic energy emitting devices, include at least a metal layer that is formed, for example, by electroplating/electroforming or casting. The metal layer includes top and bottom surfaces, and a plurality of solid integrated walls. Each of the solid integrated walls extends from the top to bottom surface and has a plurality of side surfaces. The side surfaces of the solid integrated walls are arranged to define a plurality of openings extending entirely through the layer. At least some of the walls also can include projections extending into the respective openings formed by the walls. The projections can be of various shapes and sizes, and are arranged so that a total amount of wall material intersected by a line propagating in a direction along an edge of the grid is substantially the same as another total amount of wall material intersected by another line propagating in another direction substantially parallel to the edge of the grid at any distance from the edge. Methods to fabricate these grids using copper, lead, nickel, gold, any other electroplating/electroforming materials or low melting temperature metals are described.
044951370	summary	BACKGROUND OF THE INVENTION This invention relates to nuclear reactors in general, and specifically to loop type and pool type fast breaders. More particularly, the present invention relates to a nuclear reactor of the loop type in which the space between a reactor vessel and a guard vessel disposed outside the reactor vessel is kept in a pressurized sealed state. In a pool type reactor, the space to be sealed according to the present invention is defined between a main vessel and a safety vessel disposed outside the main vessel. However, for the sake of simplicity, the terms "reactor vessel" and "guard vessel" used hereinafter include also such main vessel and safety vessel, respectively, in the pool type reactor. Conventionally, a guard vessel has been disposed outside a reactor vessel or tank. Since this guard vessel merely encompasses the reactor vessel, the space around the reactor vessel is kept at a pressure substantially equal to atmospheric pressure, and hence is kept open, in a sense. If, by any chance, damage to piping in a primary cooling system, or a core disassembly accident occurs in such a reactor, wide dispersion of radioactive substances would result. (Though every kind of safety measure has been taken in the reactor to prevent such an accident, it could be hypothetically or provisionally considered.) This problem can be solved by keeping the space between the guard vessel and the reactor vessel in a sealed state. According to the sealed arrangement, dispersion of the nuclear fuel materials, fission products and coolants to the outside can be reliably prevented even if the reactor vessel or the piping in the primary cooling system disposed near the reactor vessel is damaged. One possible method of fixing the reactor vessel and the guard vessel to each other in a sealed state is to employ bellows means between them. However, the major problem in using bellows is that bellows have a predetermined service life and must be replaced when broken. Disposing such bellows means near the reactor vessel inside a biological shield is not desirable, since a man cannot enter the biological shield when once the operation of the reactor is started. Another method of fixing the guard vessel to the reactor vessel in a sealed state would be welding. When the guard vessel is to be welded to the outside of the reactor vessel, however, it is not possible to weld all the portions to be welded from both inside and outside of the guard vessel. Namely, welding from outside all the portions of the guard vessel is of course possible, but when the inside of the guard vessel is welded, an exit portion for the welders and welding tools from the guard vessel must be left open and such exit portion cannot be welded from inside of the guard vessel. If there exists any weld portion welded only from outside, the reliability of the seal at that weld portion is lowered remarkably, thus reducing the overall reactor reliability. SUMMARY OF THE INVENTION It is an object of the present invention to provide a nuclear reactor having a high level of safety in which the space between a reactor vessel and a guard vessel is kept under a highly reliable seal. To accomplish this object, according to the present invention, there is provided a nuclear reactor in which the space between a reactor vessel and a guard vessel disposed outside the reactor vessel is kept in a sealed state. The space between the reactor vessel and the guard vessel communicates with the space outside the guard vessel through a liquid manometer structure, and the manometer structure is filled with a liquid so as to provide a liquid-sealed arrangement. In accordance with a preferred embodiment of the present invention, an inert gas or the like is charged in the sealed space between the reactor vessel and the guard vessel to keep the space in a pressurized state. This construction makes it possible to restrict the flow rate of reactor contents such as coolant or the like which flow from the damaged portion when the reactor vessel or the piping in the primary cooling system is broken, and to prevent the occurrence of an accident or escalation of the accident in case an accident does occur. In accordance with another embodiment of the present invention, a tag gas is charged in the sealed space between the reactor vessel and the guard vessel to keep the space in a pressurized state. By continuously monitoring whether or not the tag gas leaks from the sealed space, soundness of the reactor vessel, the guard vessel, the piping in the primary cooling system and the like, can be continuously inspected even during operation of the reactor. This further enhances the safety of the nuclear reactor itself. Hereinafter, preferred embodiments of the present invention will be described with reference to he accompanying drawings.
	abstract	A device for measuring gas pressure that is dependent upon the level of a liquid comprises a pressure indicator or sensor (6, 6xe2x80x2) which is connected to an upper part of a tube conduit (7, 7xe2x80x2) which is filled with gas and intended to operate in a liquid mass, said conduit having a lower part connected to an upper part of a vessel (8, 8xe2x80x2) with a volume that is larger than the volume of the tube conduit. The vessel (8, 8xe2x80x2) has an inlet (13) to let the liquid in, which inlet has a cross-sectional area that is smaller than the cross-sectional area of the vessel and that is located near the bottom of the vessel. Therefore, liquid that penetrates into the vessel via the inlet forms a surface whose level in itself in the vessel may vary depending upon the level of the surrounding main liquid mass, however normally without the liquid rising up into the tube conduit (7, 7xe2x80x2) as such.
	abstract	An X-ray optical system provides selectively a linear X-ray beam and a point X-ray beam while using an X-ray source which generates an X-ray beam having a linear section. When the point X-ray beam is selected, an X-ray intensity per unit area becomes higher. The X-ray optical system has an X-ray source, a parabolic multilayer mirror to which an aperture slit plate is attached, an optical-path selection slit device, a polycapillary optics and an exit-width restriction slit. The polycapillary optics and the exit-width restriction slit are detachably inserted into a path of a parallel beam coming from the parabolic multilayer mirror, and thus they can be removed from the path and a Soller slit and a divergence slit can be inserted instead.
050200836	summary	BACKGROUND OF THE INVENTION This Invention relates to manufacture of semiconductor chips using X-rays, and particularly to the fabrication of patterned membrane X-ray masks and their use. Forming patterned images on semiconductor chips with X-rays through X-ray masks offers a number of advantages. X-rays can form more precise images than visible or ultra-violet light because they do not suffer as much diffraction as the latter. X-ray images are thus more precise than other images and avoid the imperfections that can ruin a chip. As is known, a single imperfection can ruin a semiconductor chip, and the use of X-rays increases the statistical yields of perfect chips. X-rays also may make increased packing densities possible. Patterned X-ray membrane masks are used to form images with X-rays. In the past, such patterned X-ray membrane masks have been manufactured by depositing X-ray absorbing materials such as tantalum, gold, tungsten, etc. on the surfaces of silicon membranes. While such patterned membrane X-ray masks have proven successful for many applications, they exhibit a number of disadvantages, both in fabrication and use. OBJECTS AND SUMMARY OF THE INVENTION An object of the invention is to improve X-ray masks, their fabrication, and use. According to a feature of this invention, these objects are attained in whole or in part, by fabricating the masks so that the X-ray absorbing materials penetrate into the X-ray transparent silicon membrane. These and other features of the invention are pointed out in the claims. Other objects and advantages of the invention will become evident from the following detailed description when read in light of the accompanying drawings.
047088435	abstract	A control unit for a nuclear reactor constituted by a plurality of sealed vessels (17) communicating with the inside of the tank of the reactor (15) which they extend upwards above its cover (16); each containing a displacement mechanism for a unit absorbing neutrons in the core of the reactor. The drive motors (18) for the mechanisms are positioned in the upper part of the sealed vessels (17) within vertical aeration ducts (32). The control unit includes a supporting and insulating device for the sealed vessels (17) constituted by a vertical structure (23) fast to the cover of the tank (16), a horizontal strengthening plate (27) fixed to the upper part of the vertical structure (23) and a heat-insulating envelope (36) surrounding the sealed vessels (17) to a level below the level of the motors (18). The invention is particularly applicable to pressurized water nuclear reactors equipped with screw and nut mechanisms.
061334987	abstract	Known phosphate ceramic formulations are improved and the ability to produce iron-based phosphate ceramic systems is enabled by the addition of an oxidizing or reducing step during the acid-base reactions that form the phosphate ceramic products. The additives allow control of the rate of the acid-base reactions and concomitant heat generation. In an alternate embodiment, waste containing metal anions are stabilized in phosphate ceramic products by the addition of a reducing agent to the phosphate ceramic mixture. The reduced metal ions are more stable and/or reactive with the phosphate ions, resulting in the formation of insoluble metal species within the phosphate ceramic matrix, such that the resulting chemically bonded phosphate ceramic product has greater leach resistance.
059178744	abstract	A target includes a body having a depression in a front side for holding a sample for irradiation by a particle beam to produce a radioisotope. Cooling fins are disposed on a backside of the body opposite the depression. A foil is joined to the body front side to cover the depression and sample therein. A perforate grid is joined to the body atop the foil for supporting the foil and for transmitting the particle beam therethrough. A coolant is circulated over the fins to cool the body during the particle beam irradiation of the sample in the depression.
	description	1. Field of the Invention The invention relates to a neutron-optical component array for the specific spectral shaping of neutron beams or pulses in a neutron guide or beam hole between a fast neutron source with several moderators of different structures arranged closely adjacent each other for generating slow neutrons of different energy spectra as well as for their radiation in predetermined radiation directions and to at least one place of experiment. 2. The Prior Art Neutron beams serve in a broad spectrum of scientific examinations ranging from pure basic science to application-related examinations in the field of research of the structure of matter. Here, neutrons function quasi as sensors which penetrate into the matter. Neutrons impinging upon atoms of structured matter are either scattered in a manner characteristic of the atoms or they are absorbed by the atoms by emitting characteristic radiation. For most applications, as for instance in neutron scattering, it is necessary to provide slow neutrons which are generated by deceleration of fast neutrons obtained from nuclear reactions. Intensive neutron radiation of fast neutrons is primarily generated in research reactors either by splitting enriched uranium in a temporally constant flow or as pulses in spallation sources by crushing heavy atoms. The specific deceleration of fast neutrons is primarily carried out by so-called “moderators” which are brought into contact with the fast neutron radiation. Stated in simple terms, these are collections of matter of gaseous, liquid or solid appearance which, at a predetermined temperature, have specific characteristics. By the interaction of fat neutrons with the preferably light atoms of the moderator matter, the high energetic neutrons are strongly decelerated to the point where their energies and wavelengths are of the requisite values for experiments with condensed matter. A neutron gas of kinetic energy distribution is produced which at a given temperature may be approximated by a Maxwellian velocity distribution. This is a theoretically derived function which assigns their relative abundance to the velocities of the atoms of a gas. The effective temperature of the Maxwellian spectrum of the neutron gas is somewhat higher, however, than the temperature of the moderator matter. In this connection it is to be mentioned that neutron reflectors such as, for instance, (heavy) water, lead, beryllium, graphite, etc. also generate slow neutrons, but with a spectrum different from the spectrum which may be approximated by the Maxwell spectrum. Nevertheless, reflectors which serve primarily to increase the flow of neutrons also contribute to neutron-deceleration, so that, in a broader sense, they may, as neutron-optical components, be grouped with the moderators. Premoderators such as water and all other structures of a neutron sources capable of emitting slow neutrons may also be counted among the group of moderators. Depending upon the temperature of the moderator material, slow neutrons are differentiated between “hot”, “thermal”, and “cold” neutrons, so that the moderators may also be distinguished as “hot”, “thermal”, and “cold” moderators. In the present context, slow neutrons are those of a kinetic energy in the range of 1 eV and less. The energy of hot neutrons of higher velocity and lesser wavelength is in a range above 100 meV and are particularly suitable for scatter experiments with liquids. Thermal neutrons are of a kinetic energy in the range of between 10 meV and 100 meV, and the kinetic energy of cold neutrons lies in the range between 0.1 meV and 10 meV. Cold neutrons of relatively low velocity and large wavelength are above all of importance for applications of neutron scattering for examining biological substances. Depending on the kind of their primarily generated slow neutrons, a distinction is made between hot, thermal and cold moderators. A survey of possible moderator structures in a spallation source may be derived from paper I “Particle Transport Simulations of the Neutron Performance of Moderators of the ESS Mercury Target-Moderator-Reflection System” (downloadable from the Internet at http://www.hmi.de/bereiche/SF/ess/ESS_moderators3.pdf, state 18 January 2002). Examples thereof are the liquid hydrogen moderator with an operating temperature in the range of 25° K for generating cold neutrons and the water moderator using the ambient temperature as its operating temperature for generating thermal neutrons. However, a cold moderator also generates thermal and hot neutrons as well, and a thermal moderator also generates cold and hot neutrons, but always at a flow lower by an order of magnitude than the moderator which serves for generating primarily cold, thermal or hot neutrons. To provide the correct required neutron spectrum for different experiments with slow neutrons, the known neutron sources operate with a combination of different moderators. From Paper II “The Spallation Neutron Source Project” by Jose R. Alsonso; Proceedings of the 1999 Particle Accelerator Conference, New York, 1999, pp. 574–578, (downloadable from the Internet at http://accelconf.web.cem.ch/accelconf/p99/PAPERS/FRAL1.pdf—(State 18. January 2002), it is known to position two water moderators tempered by room temperature below the level with the target material to be crushed and two super-critical hydrogen moderators with an operating temperature of 20° K above the target plane. Each moderator exclusively provides one or more of eighteen places of experiment with the slow neutron spectrum generated by it (see FIG. 9 and Chapter 6 of Paper II). A similar structure is also known from Paper III “5.3—Material Issues for Spallation Target by GeV Proton Irradiation” by W. Watanabe (downloadable from the Internet at http://www.ndc.tokai.jaeri.go.jp/nds/proceedings/1998/watanabe_n.pdf; state 18 January 2002). It describes a target-moderator-configuration for executing high intensity and high resolution experiments with cold neutrons, in which a coupled cold moderator with a premodulator and two thermal moderators are arranged closely adjacent the target in the region of the highest and fastest neutron radiation (see Paper III, Chapter 4 (2) to (4) and FIG. 2). As an important point, the paper refers to the close proximity notwithstanding, cross-talk between the individual moderators which effects the neutron intensity, can be prevented (see Paper III, Chapter 4 (ii)). For that reason, the moderators are arranged relative to each other at such angles that their forward and rearward radiation directions or emitted neutron beams are oriented in different spatial directions without overlapping each other. In this manner, each moderator supplies about four to eight places of experiment with a neutron beam of characteristic spectrum. Moreover, reflectors are arranged between the to levels for separating the spectra. Proceeding from the known state of the art relating to the known application of moderators as described, for instance, in previously cited Paper III, it can be recognized that the provision of a neutron spectrum of slow neutrons required for a specific experiment as well as the generation thereof causes significant problems. In particular, with regard to the very complex and expensive structures of the neutron-optical components as well as the high protective measures which they require, the state of the art knows of no neutron spectrum for a single place of experiment. Each place is supplied with a neutron spectrum the maximum of which indicates the principally generated slow neutrons, from a directly associated moderator type. Changes in the spectrum of the neutron beam at a place of experiment may be realized only by significant structural changes in the structure of the moderator at extended down-times of the neutron source. Experiments in energy ranges broader than the one of a single slow neutron form are not possible or they are very inefficient. For that reason, it is an object of the invention to provide an array of neutron-optical components for the specific shaping of the spectrum of a neutron beam of the kind referred to supra which offers significant flexibility in respect of providing one neutron beam to one place of experiment, so that no extensive structural changes are required in case of change requirements. More particularly, experiments with neutrons from a larger energy range are to be made possible as well. Furthermore, the neutron beam provided by the invention is to be of high quality. The means for realizing the invention are to be simple in their structure and operation and, therefore, subject to relatively few malfunctions as well as low costs. Present aspects of safety are to be taken into consideration and additional risks are to be avoided. In the accomplishment of this object the invention provides in a neutron-optical component array for the specific shaping of neutron beams or pulses of the kind described hereinbefore for the radiation directions of the moderators to overlap directly or by further neutron-optical components in the neutron guide or at the place of experiment and for the slow neutrons of different energy spectra in an overlapping neutron beam be detected together with a multi-spectrum which is defined by the structure and number of moderators used. The energy spectra of different moderators are combined into a “multi-spectrum” by the neutron-optical component array in accordance with the invention. A neutron beam (or a neutron pulse—this alternative is always to be included when the term “neutron beam” is used) with such a multi-spectrum may be used in many different applications. As it has a broader energy spectrum than the individual neutron beams generated by a moderator, the overlapping neutron beam in accordance with the invention makes possible neutron experiments with high efficiency in a broad energy range of the impinging neutrons, e.g. between 0.1 meV and 100 meV. The composition of the multi-spectrum of the overlapping electron beam depends upon kind and number of moderators used. For instance, a cold and a thermal moderator or a cold, a thermal, and a hot moderator may be combined in their direction of propagation. In the same manner, different designs of a type of moderator may be combined to achieve a particularly broad multi-spectrum or a specially-formed multi-spectrum in terms of its emission. The combination of different modulators is limited only by structural restraints since in terms of apparatus technology the combination of the radiation direction must be realizable with a reasonable effort. In this connection, mention is to be made that other neutron-optical components present in the neutron system as well as parts of the neutron source itself may, of course, be included in the composition of the multi spectrum, with other main functions which provide for a decelerating effect on the neutrons, such as reflectors, neutron guides, and primary moderators, by combining the emitted radiation into the common neutron beam. This results in a single or multiple overlapped neutron beam for many different applications. The point of gravity of the invention resides in the combination of the individual neutron beams in a common neutron beam with a correspondingly broadened energy spectrum. Heretofore, the prior art has always proceeded from an express and deliberate separation of the effective ranges of the moderators since this seemed to be the only possibility without much effort to provide suitable slow neutron beams for yielding usable measurement results. The disadvantage of the low flexibility was accepted and corresponding numbers of places of experiment were conceived. The overlapping of the individual neutron beams from the moderators used to a common neutron beam may take place in the neutron guide as well at the place of experimenting. The first case results in the formation of a neutron beam which like a single electron beam is conducted in one neutron guide to the place of experiment and to the probe. In the second case, the different neutron beams are focused on the probe to be examined so that the overlapping neutron beam impinges directly on the probe. The advantage of this overlapping irradiation at the place of experiment itself resides in the relatively low technical complexity for combining the directions of radiation of the individual moderators. In the simplest case, the adjacent moderators are to be arrayed relative to each other at such angles that it results in a focal point of the radiation directions in the probe or slightly in front thereof. In a further development of the neutron-optical component array in accordance with the invention the radiation directions may, in case they overlap directly, be detectable at the place of experiment by a predetermined encoding scheme. In terms of the measurement results it may be important to know the different radiation directions from which the different kinds of neutrons impinge upon the probe. In a pulsed neutron source this may be carried out by monitoring the neutron flight time. In case of a it is necessary to chop the neutron beam correspondingly. Since within the slow neutrons, the cold, thermal, and hot neutron differ by the energy spectrum and, hence, by their velocity distribution, knowledge of the individual neutron flight times makes possible, on the basis of the pulses, an association to the individual moderators and, therefore, with their radiation direction relative to the probe. However, for the majority of applications in experiments it is important that all the neutron from a common spatial direction impinge upon the probe to be examined. This common spatial direction will hereafter be denominated “effective mean beam direction”. To achieve a common beam direction overlapping of the individual neutron beams by further neutron-optical components is necessary. Different components are known for the specific control of the neutron beams, all of which are suitable in the array in accordance with the invention to bring about a combination of the emissions of the moderators. Among these is the neutron guide itself which in accordance with one embodiment of the invention may on its interior surface be plated with nickel (se German patent specification DE 44 23 781 A1) and which reflects neutron impinging at predetermined especially flat angles into the interior of the tube. If two neutrons impinge the input section of the neutron guide from two different directions, for instance, they will be steered into the desired effective mean beam direction during the course of the neutron guide by the internal reflection thereof. Furthermore, in an overlapping of the radiation directions by further neutron-optical components for achieving an effective mean beam direction of the overlapping neutron beam, a further embodiment of the invention may provide for a further neutron-optical component structured as an oscillating reflector which oscillated in synchronism with a pulsed neutron source or with the chopped neutron beam of a continuous neutron source. The oscillating reflector causes the neutron beams from different moderators to be alternatingly inserted into the overlapping neutron beam with the effective mean beam direction. If, for instance, the reflector oscillates to and fro between a cold and a thermal moderator at the beat rate of a neutron pulse source and if its angle is proper in respect of the impinging cold neutrons, it will initially reflect the cold neutron pulse into the means radiation direction. Thereafter, the angle of the reflector is changed at the beat rate of the pulse so that thermal neutrons will impinge and the thermal neutron pulse is coupled in. The respective other neutron pulse will be deflected outside of the mean radiation direction. At a continuous neutron beam from a core reactor mechanical or chopper arrangements operating differently may be used for chopping the continuous neutron beam into individual pulses. In such an embodiment, measurements at the probe are to be carried at the beat rate of the neutron pulses or of the oscillator. It has already been mentioned supra that in the energy spectra of the individual moderators two marginal areas with neutron energies occur which are mainly generated by the other moderators. If in an experiment only cold neutrons have been fed to a probe, hot and thermal neutrons will nevertheless be present in the neutron beam, yet at a significantly lower quantity. In accordance with a further embodiment of the neutron-optical component array in accordance with the invention it is particularly advantageous to provide a further neutron-optical component with an energy depending switching function. In this variant of an embodiment, there is no active moving reflector switching back and forth between individual neutron beams, but a neutron-optical system is provided instead which simultaneously captures all impinging neutron. In this connection a neutron-optical component is used which is provided with an energy-selective switching function. Such components may be structured and aligned so that they pass, for instance, the central energy range of each moderator with the greatest quantity of the neutrons to be generated and couple them into the effective mean radiation direction. By contrast, they block the marginal areas with the energetically diverging neutrons. The multi spectrum of the overlapping neutron beam may be combined by the switching function by passing for the individual kinds of neutrons the corresponding neutrons from the moderators which generate them. It is thus possible for cold as well as for thermal and hot neutrons to attain a maximum neutron flow for the experiments. Neutron-optical components with an energy-selective switching function may be realized primarily by special neutron reflectors. For that reason, a further embodiment of the invention provides for the further neutron-optical component with an energy-depending switching function to be structured as a neutron reflector which continuously or intermittently passes or blocks impinging neutron by a corresponding angular alignment depending upon their energy. For further explaining the functional cooperation of the neutron reflectors, to achieve the switching action described above, reference may be had, for the sake of avoiding repetition, to the particular section of this specification. In accordance with a further embodiment of the invention, the neutron reflectors may advantageously be structured to be self-supporting or as being applied on a neutron-transparent substrate as a single layer or multi-layered reflector, with the coating being applied to one or both sides of the substrate. The multi-layered neutron reflectors are so-called “super-reflectors” with interfering properties (see German patent specification DE 198 44 300 A1). For instance, silicon and sapphire are suitable substrates. All of these neutron-optical components are of relatively simple structure and are thus inexpensive compared to other neutron-optical components. A particularly advantageous and compact structure of the invention results in accordance with another embodiment by integrating the further neutron-optical components with an energy-depending switching function into the neutron guide. As regards this embodiment, reference may be had, for the sake of avoiding repetition, to the specific portion of the description. FIG. 1 depicts the neutron-optical component array NOA for the specific spectral shaping of neutron beams or pulses. In the selected embodiment a cold moderator CNM for neutrons is arranged closely adjacent a thermal moderator TNM for neutrons. Both moderators CNM, TNM measure 12×12 cm in cross-section and are separated by a gap of 0.5 cm. Instead of a representation of an angular arrangement between the two moderators CNM, TNM their radiation directions CBL, TBL are indicated as being angular relative to each other. The cold moderator CNM emits a neutron spectrum having a maximum of cold neutrons CCN and a smaller proportion of thermal neutrons CTN. On the other hand, the thermal moderator TNM generates a maximum of thermal neutrons TTN and a lesser proportion of cold neutrons CTN. The thermal moderator TNM is arranged directly opposite a neutron guide NGT which conducts the coupled-in neutrons to a place of experiment not shown in FIG. 1. The neutron guide NGT has a cross-section of 6 cm×6 cm and extends from the neutron source also not shown in FIG. 1 by a distance of 32 m. For improving its reflective properties it is coated with nickel on its internal surface INS. By multiple flat reflection of acutely impinging neutron beams CCN, TTN it concentrates them in an effective mean radiation direction EBL to an overlapping neutron beam SBL having a multi spectrum. By attaining the effective mean radiation direction EBL, the neutrons impinge upon the probe to be analyzed quasi from one direction. The overlapping neutron beam SBL generated in the neutron guide NGT by beam overlapping has a multi spectrum of particularly high value which is composed of from the maximum ranges of the spectra only of the two moderators CNM, TNM. To obtain such a purified multi spectrum which may be used with particular advantage for experiments in a broad energy range, further neutron-optical components NOC with an energy-dependent switching function are integrated into the neutron guide NGT at its end facing the two moderators CNM, TNM at a distance of 1.5 m therefrom. In the selected embodiment, these are a simple neutron conducting super reflector RSM and a further super reflector SSM opposite the first one. They arranged at an angle of 0.72° relative to the direction of the neutron guide NGT. So that the super reflector SSM reflects or passes impinging neutrons as a function of their kinetic energy. If a different angle is selected, the other dimensions of the participating components must be changes correspondingly. Both super reflectors RSM, SSM have a length of 6.5 m and are of commercial quality m=3, i.e. their sectional angle is thrice the sectional angle of natural nickel. The super reflector SSM is applied at a thickness of 0.75 mm to a neutron transparent Si substrate. Whereas the super reflector RSM serves merely to reflect emitting neutron beams, the opposite super reflector SSM fulfills an energy and angle depending switching function. In the selected example, the super reflector SSM is constructed and set in its angle (for instance 0.72° in this example) such that it reflects the cold neutrons CCN of the cold moderator CNM into the neutron guide NGT, whereas the cold neutrons CTN from the thermal moderator TNM are reflected away from the area of the neutron guide NGT by the other side of the reflector. In the opposite case, the thermal neutrons TCN of the cold moderator CNM are guided out of the neutron guide NGT along the super reflector SSM, whereas the thermal neutrons TTN from the thermal moderator TNM may unimpededly pass through the super reflector SSM. In this manner the overlapping neutron beam SBL is composed of preferentially emitted neutrons from both moderators CNM, TNM. This ensures on the one hand that at every neutron energy switching takes place to the moderator with the higher neutron flow and, on the other hand, that the other moderator with the possibly lesser beam quality—e.g. pulse shape in case of pulsed sources—are deflected out. FIG. 2 depicts the switching function for generating the multi spectrum of the arrangement in accordance with the invention in exemplarily selected embodiment of FIG. 1. The relative transmission coefficient RTC of the entire neutron-optical system is shown as a function of the neutron wavelength NWL in nm for bother moderators CNM, TNM of FIG. 1 and may be defined as by comparison with the simple spectra in an identical neutron guide which is arranged at a distance of 1.5 m either ahead of the cold or ahead of the thermal moderator CNM, TNM. If neutron energy greater than 20 meV (this corresponds to a neutron velocity in excess of 2,000 m/sec or, by way of equivalence, to a neutron wavelength below 0.2 nm) are needed in an experiment, thermal neutrons TTN exclusively will be available in the combined multi spectrum. At neutron energies less than 5 meV (corresponding to a neutron velocity of less than 1,000 m/sec or, by way of equivalent, to a neutron wavelength of more than 0.4 nm) the supply of neutrons is satisfied with cold neutrons CCN almost exclusively from the cold moderator CNM. In a transitional range between 5 meV and 20 meV the neutrons TTN, CCN are fed in the overlapping neutron beam SBL to the experiment from both moderators TNM, CNM as a mixture with different proportions.
	summary
	summary
	abstract	A bowtie filter is constructed to have a fluidic envelope filled with attenuating fluid and a displacement insert that can present various x-ray attenuation profiles during a scan. The insert is designed to displace the attenuating fluid to achieve a denied attenuating or filtering profile. The insert can be rotated, twisted, moved, and otherwise contorted within the fluidic envelope as needed during the course of a scan. As the angle, position and shape of the zombie is changed, the x-ray profile of the filter changes. The insert may have a default shape when at rest, but can have its shape changed when external forces are placed thereon. As x-ray filtering needs change during the course of the scan, the insert can be compressed, stretched, and/or contorted to achieve additional filtering profiles.
052788776	summary	BACKGROUND OF THE INVENTION The invention concerns a process for dismantling buried unsheltered equipment at risk of radioactive contamination, particularly the kind that extends in a straight line, such as shut-down effluent pipes. As part of a general radioactive-contamination risk protection policy, waste and equipment that may be contaminated must be eliminated when the equipment is taken off line and the areas it occupied must be made available for other uses. Indeed, even when pipes are buried in the ground, they may eventually deteriorate and spread contamination by infiltration and/or irrigation, which is very difficult to contain. Generally, a state of noncontamination or decontamination means a state in which the radioactivity of a material is below a predetermined threshold, for example, one stipulated by law. SUMMARY OF THE INVENTION The invention proposes a dismantling process that meets the requirements of strictness and confinement essential for work in an area potentially subject to radioactivity. More precisely, the invention proposes a process for dismantling unsheltered, buried equipment which is at risk of contamination and is possibly irradiating, characterized by the fact that sections of the equipment are dismantled in units, under the cover of a movable vessel placed above the section to be dismantled. According to one special method of implementing the process in the invention, after the vessel is put into place, the section to be dismantled is first: surrounded by a protective envelope; then PA1 disengaged on site in its trench; PA1 separated from the rest of the equipment; PA1 put into the vessel in pieces; PA1 placed in a container that is in the vessel; and PA1 evacuated from the vessel in the container. PA1 It allows the vessel to be positioned with great precision on spaces or land with reduced access and/or on land where there has been an accident; PA1 It allows work to be done in the ditch without risk of landslide; and PA1 It allows vessel supports to be set up quickly and inexpensively. PA1 a self-supporting structure built around a deck that has a central opening; this structure supports various mobile lifting or handling tools and an outside envelope equipped with a handling room and/or at least one window with a movable cover; PA1 a flexible protective curtain located around the central opening to surround the work area where the section being dismantled is located and to isolate it from the outside, leaving the area accessible from the vessel; PA1 a forced ventilation system with a filter to filter out contaminated dust from inside the vessel. Moreover, depending on the radioactivity of the debris and the ditch where the dismantled section was located, any soil whose radioactivity exceeds a predetermined threshold is evacuated from the ditch to an appropriate storage area. One option is terracing to disengage ahead of time the sections to be dismantled after the section presently being dismantled, during which any waste whose radioactivity is higher than a predetermined threshold is stored temporarily in said vessel before it is evacuated to the storage area. In one variation of the process in the invention, supports for the movable vessel, such as blocks or a support rail are set up on either side of the equipment as the equipment is being dismantled. The vessel is then moved, for example, by crane. This option of setting up temporary supports such as blocks is very advantageous in several respects: The invention also concerns a vessel for implementing various versions of the process described above; the vessel contains: According to one variation of the vessel that is consistent with the invention, the vessel also includes a movable internal cover over the central opening to isolate the inside of the vessel for a period of time from the work area where the section being dismantled is located. Advantageously, the vessel also includes a moving-head-type vacuum device with a filter to pick up dust in that area. In one special embodiment of this variation, the forced ventilation device and the vacuum ventilation device are made so that once the movable inside cover is put in place, the work area is depressed in relation to the inside of the vessel and the inside of the vessel itself is depressed in relation to the outside air. According to another variation of the vessel that is consistent with the invention, the flexible curtain is composed of a bellows, one end of which is joined to the edges of the central opening and the other end of which bears a rigid framework for housing the bellows when it is retracted. Advantageously, this bellows is maneuvered in extended or retracted position by a lifting device with a winch on the frame. According to yet another variation of the vessel that is consistent with the invention, two sides of the opening form a rolling path for a cart suspended in the opening. Advantageously, the cart carries a hydraulic scoop. According to another embodiment of the vessel that is consistent with the invention, the deck has an area for a container accessible from a window with a movable cover located preferably in the part forming the roof of the vessel and, possibly, a work area for the section being dismantled. Advantageously, the deck also has an area for a container of soil accessible from another window with a movable cover located preferably in the part forming the roof of the vessel. As an option, the vessel may include a hopper equipped with a controlled-opening helmet that passes through the outside envelope and comes out above the area for the soil container. This hopper takes the contaminated terracing soil extracted by a hydraulic scoop with closed buckets located on the outside of the vessel. Such an arrangement makes it possible to confine debris from the equipment being dismantled, on one hand, and soil from the ditch and the embankment that may be contaminated. According to yet another variation of the vessel consistent with the invention, the vessel is constructed around two half-shells assembled so that they can be dismantled on both sides of a vertical median plane, preferably parallel to the largest side of the deck. According to yet another variation of the vessel consistent with the invention which can be used particularly for dismantling equipment with straight extensions, the central opening is rectangular and is adapted so it can be positioned carefully and aligned with the specific section to be dismantled.
	abstract	Provided herein are charge generating devices and methods of making and use thereof. The charge generating devices comprise a substrate having a top surface; a plurality of spaced-apart three-dimensional elements disposed on the top surface of the substrate; and a plurality of cavities formed by the plurality of spaced-apart three-dimensional elements, the plurality of cavities being the area between the plurality of spaced-apart three-dimensional elements. The charge generating devices can further comprise a radioactive layer disposed on at least a portion of the plurality of spaced-apart three-dimensional elements and the top surface such that the plurality of cavities and the top surface are substantially coated by the radioactive layer. In some examples, the charge generating devices can comprise a radiation material and/or a scintillating material disposed within at least a portion of the plurality of cavities.
	abstract	The invention concerns a spacer (14) for a nuclear boiling water reactor. The spacer (14) comprises a plurality of cells (16) for holding or allowing elongated elements (12) to pass through the cells. Between the cells (16) there are a plurality of flow channels (18). The spacer comprises at least a plurality of deflecting members (22). The deflecting member comprises a vane (24) which extends in a direction from a cell (16) into the neighbouring flow channel (18). The vane is inclined relative to a vertical plane (26) and is wider in its upper part than in its lower part. The invention also concerns a fuel assembly for a nuclear boiling water reactor, comprising a deflecting member with vane of similar construction.
	abstract	The use of hydroxyiminoalkanoic acids including at least four carbon atoms as anti-nitrous agents in operations of reductive stripping of plutonium. The invention may be useful in any method for processing spent nuclear fuels that includes one or more operations of reductive stripping of plutonium and, more particularly, in the PUREX method as implemented in modern nuclear fuel processing plants, as well as in processes derived therefrom.
047042460	summary
	summary
	description	The present invention relates to a repairing method of a cylindrical body such as a control rod drive mechanism (hereinafter referred to as “CRD”) provided in a reactor bottom of a reactor vessel of a boiling water reactor and a pressurized-water reactor and such as a penetrating (through) tube for a reactor core measuring sensor provided in the reactor bottom of the pressurized-water reactor. The present invention particularly, relates to a reactor bottom repairing method for preventing reactor water from leaking from a crack occurring in a vicinity of a welded portion of the cylindrical body to thereby ensure the repair or maintenance and preventive repair or maintenance for preventing reactor water from leaking in the future. FIG. 6 is an enlarged view of a partial cross section illustrating an entire configuration of a boiling water reactor. FIG. 7 is an enlarged sectional view of a reactor bottom portion shown in FIG. 6. As shown in FIGS. 6 and 7, a CRD housing 3 for storing a CRD and a stub tube 2 for supporting the CRD housing 3 are provided for a bottom head 1 of a reactor pressure vessel in a boiling water reactor. As illustrated in FIG. 7, the stub tube 2 is welded to a reactor pressure vessel bottom head (or lower-side end plate) (hereinafter simply referred to as a “bottom head”) 1 by means of a welded portion 4. The CRD housing 3 passing through the stub tube 2 and the bottom head 1 is welded to an upper portion of the stub tube 2. Further, as shown in circular portions “a”, “b” and “c” of FIG. 7, sensitivity to stress corrosion cracking becomes high particularly in a heat affected portion in a vicinity of the welded portion. It has been reported that in older plants built before knowledge of stress corrosion cracking was available, cracking may occur due to a fault inherent in the welded portion, and as a result reactor water leaks through the cracking. In a conventional method, when cracking occurs in a stub tube in a penetrating portion of a CRD housing in the reactor bottom portion and a reactor water leakage is detected, a method or treatment replaces an entire CRD housing and stub tube or expands and deforms the tube of the CRD housing from inside by a roller to be pressed against a through hole of the bottom head of the pressure vessel to plug the leakage route. Alternatively, the area with a leak can be enclosed with a cylindrical mechanical seal member. As the repairing method in a case of a shallow and non-penetrating crack, a method in which a cracked portion is removed by mechanical processing or electrical-discharge machining and then overlay repairing is performed by TIG welding or laser welding, or a welding method for welding only a surface crack, has been studied. The conventional method of replacing the entire CRD housing and stub tube requires a long time for preparatory and replacement work, resulting in a long time needed for stopping of the operation of the reactor, and hence, highly increasing working cost. Although the roller tube expansion method is a simple and easy method, re-leakage from a tube expansion portion is likely to occur due to thermal deformation in the operating cycle of the reactor and thus, complete sealing for a long period of time is difficult. Further, the seal method of attaching a sealing mechanism member is a method only applicable to a limited leak position. According to the conventional methods (see Patent Documents 1 to 6) in which the cracked portion is removed by mechanical processing or electrical-discharge machining and then overlay repairing is performed by TIG welding or laser welding, there is a high possibility that a new crack occurs due to a flaw of an existing stub tube or a welded portion thereof. It is difficult to perform a particular treatment on this portion and thus the conventional methods are not permanent measures. Furthermore, in the conventional technology, generation of a crack and the repairing thereof in a cylindrical body such as a penetrating tube for a reactor core measuring sensor provided in the reactor bottom portion of the pressurized-water reactor constitute a significant matter to be solved. Patent Document 1: Japanese Patent Laid-Open No. H10-030991 Patent Document 2: Japanese Patent Laid-Open No. 2003-320472 Patent Document 3: Japanese Patent Laid-Open No. 2004-226329 Patent Document 4: Japanese Patent Laid-Open No. 2004-294372 Patent Document 5: Japanese Patent Laid-Open No. 2006-337175 Patent Document 6: Japanese Patent Laid-Open No. 2008-020447 As described above, according to the conventional techniques, since it takes a long period of time to perform preparatory and replacement work, the operation of the reactor is stopped for a long period of time, and high costs for completing the working is involved, and the re-leakage from a tube expansion portion may occur due to thermal deformation in the operating cycle of the reactor. Thus, it is difficult to perform sealing for a long period of time, and furthermore, in the seal method of attaching a sealing mechanism member, the leaking position is limited and a new crack may occur, resulting in not providing a permanent countermeasure. As a permanent method, although there is provided a method of replacing the entire cylindrical body provided so as to pass through the reactor vessel including the CRD housing and the stub tube, it is difficult to apply such method to a case in which leaking occurs from many portions of the stub tubes. In view of such circumstances, the present invention has been conceived, and an object of the present invention is to provide a reactor bottom repairing method as a permanent measure against leakage without replacing a cylindrical body provided so as to pass through a reactor vessel including an existing CRD housing and a stub tube. In order to achieve the aforementioned object, the present invention provides a reactor bottom portion repairing method, which is a method of sealing a crack occurring on a surface of a cylindrical body passing through and fixed to a reactor bottom portion, the method comprising: emitting a heating laser beam to a cracked part to remove moisture from the cracked part; subsequently emitting a welding laser beam to the cracked part to heat and melt the cracked part; and emitting the heating laser beam and the welding laser beam to a crack of an entire surface of the cylindrical body inside the reactor and a welded portion between the cylindrical body and the reactor bottom portion to thereby prevent a new crack from occurring. Further, the present invention provides a reactor bottom portion repairing method which is a method of sealing a crack occurring on a surface of a stub tube in a CRD housing penetrating portion in a reactor bottom portion of a reactor pressure vessel of a boiling water reactor, the method comprising: removing moisture from inside the crack by emitting a heating laser beam; subsequently heating and melting the crack by emitting a welding laser beam; and emitting the heating laser beam and the welding laser beam to the entire stub tube and the welded portion between the stub tube and a bottom head to apply overlay welding to the surface with a material having a low stress corrosion cracking sensitivity and to repair the crack on the stub tube surface and the crack on the surface between the bottom head and the welded portion for preventing a new crack from occurring. Further, the present invention provides a reactor bottom portion repairing method, the method comprising: covering a CRD housing with a cap-shaped member so as to cover from an upper end of the CRD housing up to an upper end surface of the stub tube, the cap-shaped member being made of a material having a shape not interfering with an overlay welded portion at a corner between the CRD housing and the stub tube thereinside, having an outside diameter substantially matching that of the stub tube, and having a low stress corrosion cracking sensitivity; applying laser seal welding to the stub tube in a lower end portion of the member; and applying laser seal welding to a side surface of the CRD housing in an upper portion thereof to thereby prevent reactor water from leaking from the welded portion between the CRD housing and the stub tube and a heat affected portion of the CRD housing. The present invention emits a heating laser beam to a cracked part on a surface of the cylindrical body provided to pass through the reactor bottom portion to remove moisture from the cracked part; subsequently emits a welding laser beam to the cracked part to heat and melt the cracked part; and emits the heating laser beam and the welding laser beam to an entire surface of the cylindrical body inside the reactor and a crack of the welded portion between the cylindrical body and the reactor bottom portion to thereby prevent a new crack from occurring. Hereinafter, with reference to the accompanying drawings, the description will focus, according to the present invention, on an embodiment of a preventive repairing or maintenance method of repairing a crack on a surface of a welded portion between a stub tube surface and a bottom head and preventing a new crack from occurring, and on an embodiment of a method of attaching and welding a cap-shaped member for preventing reactor water from leaking even in a case of a crack occurring and penetrating through an overlay welded portion at a corner of a CRD housing or a heat affected portion. FIG. 1 is an enlarged view of an essential portion for explaining a reactor bottom repairing method according to a first embodiment of the present invention. FIGS. 2 (a), (b), and (c) are views explaining operations for procedure. This embodiment represents a method of sealing a crack occurring on a surface of a stub tube 2 in a penetrating portion of a CRD housing through a bottom head 1 in a reactor bottom portion of a reactor pressure vessel of a boiling water reactor (BWR). In this case, a heating laser beam is emitted to remove moisture inside the crack. Then, a welding laser beam is emitted to heat and melt the crack. The heating laser beam and the welding laser beam are emitted to the entire surface of the stub tube 2 and a weld portion (which may be read selectively hereinafter as a portion to be welded or a welded portion) between the stub tube 2 and the bottom head 1 to thereby perform an overlay welding to the surface with a material having a low stress corrosion cracking sensitivity to repair the crack on the surface of the stub tube 2 and the surface of the weld portion between the stub tube 2 and the bottom head 1 for preventing a new crack from occurring. It is to be noted that the inventors of the subject application have already proposed an apparatus having six control shafts to which the welding method is accessibly applied to around the stub tube. More specifically, the six control shafts include: a pivot (swivel) shaft of the entire apparatus; an advance shaft for driving a welding torch in a radial direction; a vertical shaft for driving the welding torch in a vertical direction; an inclined drive shaft for driving the welding torch along an inclined angle between an inclined surface of a bottom head and a stub tube; a torch rotating shaft for continuously changing the torch direction from the stub tube side surface up to the bottom head; and a head rotating shaft for correcting the torch position shifted due to the rotation of the inclined drive shaft. In addition, there is provided a mechanism for driving the inclined drive shaft in an arc shape changing with an inclined angle of “0” (zero) on the valley side, a maximum inclined angle at an intermediate portion, and an inclined angle of “0” on the mountain side according to the pivot angle of the entire apparatus so as to follow the welded portion between the stub tube and the bottom head changing three dimensionally therearound. Hereunder, there will be mentioned a welding method of repairing the crack in the stub tube or preventing a new crack from occurring by using the aforementioned welder to apply overlay welding to a weld portion between the existing surface of the stub tube 2 and the bottom head 1 with a material having a low stress corrosion cracking sensitivity. As illustrated in FIG. 1, according to the present embodiment, a welded portion 4 is formed by generating a seal weld layer on the weld portion 4 between the surface of the stub tube 2 and the bottom head 1. Then, as illustrated in FIGS. 2 (a), (b), and (c), there is executed a method of preventing reactor water from leaking from a crack formed in a welding heat affected portion. First, a heating and welding laser beam welder is used to apply a seal weld 6 to the weld portion 4 between (FIG. 1) the bottom head and the lower portion of the stub tube 2 along an inclined portion of the bottom head 1. In this case, the welder has a structure in which the attachment opening of the laser fiber is located on the mountain side to protect the interference with the bottom head 1. Thus, an apparatus performing different handlings depending on the left or right direction is used to weld an inclined portion on the right side and an inclined portion on the left side. Further, a cladding layer 1a is formed on an inner surface of the bottom head 1. Then, a heating and welding laser beam welder is used to apply a seal weld 7 horizontally to a side surface portion of the stub tube 2 for performing a filling process. At this time, a way how to use and handle the welder is not particularly limited. The upper end portion of the stub tube is angled, and thus, the filler metal position cannot be fixed. Therefore, though slightly, a non-weldable portion 7a will remain. Considering the dilution with the existing base material or weld metal, the seal weld layers 6 and 7 are made of three layers with a thickness of approximately 3 mm, but the seal weld layer may consist of one layer simply for the purpose of plugging the leak portion. Further, in a general welding method, there occurs an event of emitting water moisture remaining in a crack, whereas in the welding method mentioned above, it becomes possible to perform a sound seal weld layer without a pit due to the moisture emitting event. FIGS. 3 (a), (b), and (c) are views showing further aspects of the present invention, and more specifically, these views show a method including procedures and operations for applying a seal weld in a state covered with a cap-shaped member 8 to thereby prevent reactor water from leaking through a crack in a welded portion 5 between the CRD housing and the stub tube and the heat affected portion of the CRD housing shown in FIG. 1. FIGS. 3 (a), (b), and (c) represent procedures following the generation of the seal weld layers 6 and 7 on the weld portion between the surface of the stub tube 2 and the bottom head 1 according to the first embodiment shown in FIG. 2. First, as illustrated in FIG. 3(a), the CRD housing 3 is covered with a cap-shaped member 8 so as to cover the CRD housing 3 from an upper end of the CRD housing up to an upper end surface of the stub tube. The cap-shaped member 8 is made of a material having a shape not interfering with an overlay welded portion 5 at a corner between the CRD housing 3 and the stub tube 2 (which is behind the seal weld layer 7), having an outer diameter substantially matching with that of the stub tube 2, and having a low stress corrosion cracking sensitivity. Then, as illustrated in FIG. 3(b), spot laser welding is applied to the stub tube 2 so as not to move the lower end of the cap-shaped member 8 by applying the laser seal welding 9 to the stub tube and the lower end portion of the cap-shaped member 8 including the non-welded portion 7a in the upper portion of the stub tube 2 not welded in the first embodiment. At this time, preceding the application of the welding beam, an inert gas is sprayed over the welding surface to remove moisture in a gap 9a between the lower surface of the cap-shaped member 8 and the upper end surface of the stub tube, thereby providing a sound seal welded portion 9 without a pit due to the moisture emitting event by suppressing water evaporation. Finally, as illustrated in FIG. 3(c), welding is applied to a laser seal weld portion 10 between the upper portion of the cap-shaped member 8 and the side surface of the CRD housing 3. At this time, in advance of applying the welding beam, an inert gas is sprayed over the weld portion of the CRD housing to a lower water level of an inner portion 8a of the cap-shaped member. The moisture is thereby removed from the melt portion and the moisture emitting event is prevented. In the above procedure, although the unremoved moisture and inert gas remain in the inner portion 8a of the cap-shaped member, but that does not cause a problem. According to the present embodiment, the welded portion between the stub tube and the bottom head 1 is welded as the seal welded portion 6, and the side surface of the stub tube is welded as the seal welded portion 7. Subsequently, the cap-shaped member 8 covering the welded portion between the CRD housing and the stub tube is joined by the seal welded portion 10 welded to the CRD housing and the seal welded portion 9 welded to the upper portion of the stub tube. According to the present embodiment, there is provided a preventive repairing method comprising: covering a CRD housing with a cap-shaped member so as to cover the CRD housing from an upper end thereof up to an upper end surface of the stub tube, the cap-shaped member being made of a material having a shape not interfering with an overlay welded portion at a corner between the CRD housing and the stub tube thereinside, having an outside diameter substantially matching that of the stub tube, and having a low stress corrosion cracking sensitivity; applying laser seal welding to the stub tube in a lower end portion of the member; and applying laser seal welding to a side surface of the CRD housing in an upper portion thereof to thereby prevent reactor water from leaking even in a case of a crack occurring and penetrating through an overlay welded portion at a corner between the CRD housing and the stub tube or a heat affected portion of the CRD housing. FIGS. 4 and 5 illustrate an example of a pressurized-water reactor (PWR) according to further aspects of the present application. FIG. 4 is a longitudinal sectional view illustrating the entire configuration of a reactor vessel 12 of a pressurized-water reactor 11. As illustrated in FIG. 4, a reactor core 13 is provided inside the reactor vessel 12. An instrumentation tube 14 extends downward from the reactor core 13. The instrumentation tube 14 passes through a plurality of penetrating tubes 15 provided in a reactor bottom portion 12a and extends outside the reactor bottom portion. In other word, the instrumentation tube 14 extends from the lower portion of the reactor core 13, passing through inside the penetrating tube 15, and reaches an upper portion of the reactor core 13 in which a nuclear fission reaction occurs. An unillustrated moving sensor is inserted into the instrumentation tube 14 to obtain a combustion state of the reactor core 13. Each penetrating tube 15 is connected to a measuring drive apparatus 17 through a wiring 16, and a measurement signal from a sensor is measured by the measuring drive apparatus 17 provided outside the reactor vessel 12 and then transferred to a control panel 18 so as to be used for reactor operation. FIG. 5 is an enlarged longitudinal sectional view of the penetrating tube 15 in the bottom portion (shown with the capital “A”) of the reactor vessel 12 shown in FIG. 4. As illustrated in FIG. 5, the penetrating tube 15 is provided so as to pass through a bottom portion 12a of the reactor vessel 12. The penetrating tube 15 holding the instrumentation tube 14 vertically passes through a hole 20 of the reactor bottom portion 12a and is welded to an inner surface of the reactor bottom portion 12a. Further, a seal 21 is interposed between the instrumentation tube 14 and the penetrating tube 15. The seal 21 prevents water from leaking from inside the reactor. Thus, according to the present embodiment, the instrumentation tube 14 is held inside the penetrating tube 15 which is a cylindrical body penetrating through and fixed to the reactor bottom portion 12a. More specifically, a heating laser beam is emitted to a cracked portion of the penetrating tube 15 to remove moisture from inside the crack. Then, a welding laser beam is emitted to the cracked portion to heat and melt the cracked portion (welded portion 19). In the welded portion 19, the heating laser beam and the welding laser beam are emitted to a crack on the entire surface of the penetrating tube 15 inside the reactor and the welded portion 19 between the penetrating tube 15 and the reactor bottom portion 12a. The method mentioned hereinabove allows the cracked portion to be heated and melted by welding laser beam irradiation, and effectively prevents a new crack from occurring. 1 . . . reactor pressure vessel bottom head, 1a . . . cladding layer in inner surface of bottom head, 2 . . . stub tube, 3 . . . CRD housing, 4 . . . welded portion between stub tube and bottom head, 5 . . . welded portion between stub tube and CRD housing, 6 . . . seal weld portion between stub tube and bottom head, 7 . . . seal welded portion welded to side surface of stub tube, 7a . . . portion not seal-welded to side surface of stub tube, 8 . . . cap-shaped member, 8a . . . space enclosed with cap-shaped member, 9 . . . seal weld portion between cap-shaped member and upper portion of stub tube, 9a . . . gap between cap-shaped member and upper portion of stub tube, 10 . . . seal weld portion between cap-shaped member and CRD housing, 11 . . . pressurized water reactor, 12 . . . reactor vessel, 12a . . . bottom portion of the reactor vessel, 13 . . . reactor core, 14 . . . instrumentation tube, 15 . . . penetrating tube, 16 . . . wiring, 17 . . . measuring drive apparatus, 18 . . . control panel, 19 . . . welded portion, 21 . . . seal
043371183	description	DESCRIPTION OF THE PREFERRED EMBODIMENT In FIG. 1, there is shown a nuclear power plant including a boiling-water reactor (BWR) 2 provided with a core, not shown, in a pressure vessel 4. A plurality of control rods, not shown, are selectively moved into and out of the core by a rod drive system 6. Contained in the pressure vessel 4 is a coolant (light water) which is recirculated through the core by recirculation pumps 8 which receive part of the coolant and forces it to flow into jet pumps within the pressure vessel so that the coolant flows upward through fuel assemblies in the core. The heat produced by the fuel assemblies is transferred to the coolant and a head of steam is produced in the upper portion of the pressure vessel 4. The steam is supplied to a turbine 10 which drives an electrical generator 12. The turbine 10 exhausts to a condenser 14 and the resulting condensate is returned as feedwater to the pressure vessel 4 through conduit means, not shown. Located on the discharge side of each recirculation pump 8 is a control valve 16 having its opening varied by a flow control system 18, to adjust the recirculation flow rate of coolant and thus control the core coolant flow rate. Alternatively, control of the core coolant flow rate may be effected by controlling the number of revolutions of the recirculation pumps 8. The flow control system 18 will be described in detail. A reactor power change demand signal is applied to a main controller 22 either manually or as a load speed deflection signal from a turbine control mechanism 20. A neutron flux controller 26 produces a flow rate demand signal as a function of the difference between an output signal of the main controller 22 and a detected value signal of a neutron monitoring system 24. A flow rate controller 30 supplies a signal through a function generator 32 to hydraulic control means 34 for the control valves 16 so as to bring the difference between the output signal of the neutron flux controller 26 and the detected value signal of a recirculation flow rate measurement system 28 to nil. The openings of the control valves 16 are adjusted by this signal to thereby control the recirculation flow rate and thus the core flow rate to a demanded level. Control is effected by a similar system when control of the recirculation flow rate is effected by adjusting the number of revolutions of the recirculation pump 8. 36 is a turbine bypass valve. One example of the operation plan of a BWR which may be practiced by controlling the core coolant flow rate by this flow control system 18 will be described by referring to FIG. 2. In FIG. 2, the abscissa represents the core flow rate, and the ordinate indicates the reactor power level. As aforesaid, although the core flow does not show the same rate as the recirculation flow, there is a uniform relation between them. Thus, it will be noted that the core flow rate and the recirculation flow rate can be substituted for each other, and the core flow rate can be detected by the recirculation flow rate measurement system 28. At initial stages of operation of the reactor, the reactor is operated at point A on a flow rate control curve 38. With the lapse of time, a reduction is caused in reactivity owing to fuel consumption, resulting in a fall in reactor power. To compensate for this reduction in reactor power, the core flow rate is increased to maintain the reactor power at a high level by utilizing a change in the manner in which voids are formed in the reactor. By gradually increasing the core flow rate in this way, it is possibel to maintain the reactor power at a desired level for about one to two months. After the core flow rate has reached 100% or when operation of the reactor is performed at point B, no further increase in flow rate is permissible, so that the core flow rate is temporarily reduced to move the reactor operating point to C at which reactor power is reduced. By changing the control rod insertion ratio at point C, the reactor operating point moves from C to D, and then returns to point A following an increase in the core flow rate. Thus the reactor is operated in a cycle lasting one to two months. In order to prevent an excessive rise in reactor power which might result from deviation of the operation of the reactor from the aforesaid plan during operation of the plant, there is provided a core monitoring system including an average power range monitor (APRM) 40, a thermal power monitor (TPM) 42 and a rod block monitor (RBM) 44, which receive signals from the neutron monitoring system 24 and recirculation flow rate measurement system 28 and transmits a rod block signal or scram signal to the rod drive system 6. The core monitoring system further includes an operating region monitor (ORM) 46 which is operative, when an excessive rise in reactor power is caused particularly by an increase in the core flow rate, to block the increase in the core flow rate or run-back the flow rate upon the power level reaching a predetermined threshold short of the scram threshold. More specifically, ORM 46 receives signals from the neutron monitoring system 24 and recirculation flow rate measurement system 28 and transmits a coolant block signal or run-back signal to the flow controller 32 of the flow control system 18 subsequently to be described. APRM 40 will now be described by referring to FIG. 3. APRM 40 includes an averaging circuit 48 for receiving signals from the neutron monitoring system 24 including a plurality of local power range monitors (LPRMs) and averaging these signals to produce the power level of the reactor. The signal from the averaging circuit 48 is transmitted to a comparator 50. Meanwhile a rod block threshold circuit 52 is set beforehand at a power level of rod block threshold as a function of the core coolant flow rate as shown at a line 54 in FIG. 2. The rod block threshold circuit 52 receives a signal from the recirculation flow rate measurement system 28 and transmits to the comparator 50 a threshold level signal corresponding to the prevailing flow rate. Upon receiving these signals from the two circuits 48 and 52, the comparator 50 compares them and transmits a comparison signal to a signal generator 55 which transmits, when the power level is higher than the threshold level, a rod block signal to the rod drive system 6. The signal from the averaging circuit 48 is also transmitted to another comparator 56 which also receives a signal from a scram threshold circuit 58. The scram threshold circuit 58 is set beforehand at a power level of scram threshold as shown at a line 60 in FIG. 2. The second comparator 56 compares the signals from the circuits 48 and 58 and transmits a comparison signal to the signal generator 55 which transmits, when the power level is higher than the threshold level, a scram signal to the rod drive system 6. Thus, APRM 40 monitors a rise in the power level of the reactor transmits a rod block signal to the drive system 6 when the power level has reached the rod block threshold line 54 shown in FIG. 2, to thereby block control rod withdrawing. For example, when the power level reaches about 106% of the rated power in a rated power operation, control rod withdrawing is blocked. Also, APRM 40 monitors the power level of the reactor which might be caused primarily by control rod withdrawing, an increase in flow rate and a rise in the pressure in the pressure vessel 4 caused by shutoff of the load or the like. When this power level reaches the scram threshold line 60 shown in FIG. 2, APRM 40 transmits a scram signal to the rod drive system 6 to scram the reactor. Scramming takes place when the power level reaches about 120% of the rated power, for example. TPM 42 will now be described by referring to FIG. 4. Like APRM 40, TPM 42 includes an averaging circuit 62 for receiving signals from LPRMs of the neutron monitoring system 24 and averaging local power levels to produce the power level of the reactor. The averaging circuit 62 supplies a signal to a time delay circuit 64 for conversion to a thermal power level. The delay circuit 64 transmits a signal to a comparator 66 to which a signal from a scram threshold circuit 68 is also supplied. The scram threshold circuit 68 is set beforehand at a power level of scram threshold as a function of the core coolant flow rate as indicated by a line 70 in FIG. 2, for example, and transmits to the comparator 66 a threshold level signal corresponding to the prevailing core coolant flow rate upon receipt of a signal from the recirculation flow rate measurement system 28. The comparator 66 compares these two signals from the circuits 64 and 68 and transmits a comparison signal to a signal generator 72 which transmits, when the thermal power level is higher than the threshold level, a scram signal to the rod drive system 6. Thus, TPM 42 monitors a rise in the thermal power level which might be cause primarily by control rod withdrawing and a rise in the flow rate, and supplies a scram signal to the rod drive system 6 when the thermal power level has reached the scram threshold line 70 shown in FIG. 2, thereby scramming the reactor. The reactor is scrammed in rated power operation when the thermal level reaches about 115% of the rated power, for example. RBM 44 will now be described by referring to FIG. 5. RBM 44 includes an LPRMs signal selecting circuit 74 for receiving signals from LPRMs of the neutron monitoring system 24 for selection of these signals. The circuit 74 supplies a signal to a comparator 76 to which a signal from a rod block threshold circuit 78 is also supplied. The rod block threshold circuit 78 is set at a power level of rod block threshold beforehand as a function of the core coolant flow rate and transmits to the comparator 76 a threshold level signal corresponding to the prevailing core coolant flow rate upon receipt of a signal from the recirculation flow rate measurement system 28. The power level of rod block threshold at which the circuit 78 is set is not shown in FIG. 2. However, the power level is generally below the line 54 by about 1-3%. The comparator 76 compares the signals from the two circuits 74 and 78 and transmits a comparison signal to a signal generator 80 which transmits, when the selected local power level is higher than the threshold level, a rod block signal to the rod drive system 6. Thus, RBM 44 monitors a rise in the local power level which might be caused by control rod withdrawing and transmits, when the local power level reaches the rod block threshold set beforehand, a rod block signal to the rod drive system 6 to block control rod withdrawing. The nuclear reactor continues its operation even if the control rod withdrawing is blocked by APRM 40 or RBM 44. It is possible to operate again the blocked control rods if other control rods are inserted or the core coolant flow rate is reduced to thereby reduce the power level. Last but not the least important is an operating region monitor (ORM) 46 which constitutes the characterizing feature of the present invention. Referring to FIG. 6, ORM 46 includes an averaging circuit 82 for receiving signals from LPRMs of the neutron monitoring system 24 and averaging the local power levels to produce the power level of the reactor. The averaging circuit 82 transmits a signal to a comparator 84. ORM 46 also comprises a coolant block threshold circuit 86 for receiving a signal from the recirculation flow rate measurement system 28. The circuit 86 is set at a power level of coolant block threshold determined as a function of the core coolant flow rate as indicated by a line 88 in FIG. 2, and transmits to the comparator 84 a threshold level signal corresponding to the prevailing core coolant flow rate upon receipt of a signal from the system 28. The comparator 84 compares the two signals from the circuits 82 and 86 supplies a comparison signal to a signal generator 90 which transmits, when the power level is higher than the threshold level, a coolant block signal to the flow rate controller 30 of the flow control system 18. The flow rate controller 30 adjusts the openings of the control valves 16 through the function generator 32 and hydraulic control means 34 so as to block the increase in the recirculation flow rate and thus the increase in the core coolant flow rate, thereby maintaining the core flow rate at the blocked level. Thereafter, the core flow rate is manually returned to a normal operating region 92 as shown in FIG. 2. Alternatively, the signal generator 90 may be modified to generate a coolant run-back signal. In this case, the flow controller 30 which receives the coolant run-back signal adjusts the openings of the control valves 16 through the function generator 32 and hydraulic control means so as to automatically run-back or reduce the recirculation flow rate and thus the core coolant flow rate to a minimum rate. The signal generator 90 may be further modified to selectively produce a coolant block signal or a coolant run-back signal. Thus, ORM 46 monitors the reactor power level which might be caused by an increase in the core coolant flow rate. When the reactor power level reaches a coolant block threshold line 88 shown in FIG. 2, ORM 46 transmits a coolant block signal or a coolant runback signal to the flow control system 18, to thereby block the increase in the core coolant flow rate or to thereby run-back the core coolant flow rate. Thus an excessive increase in reactor power which might otherwise be caused by an increase in the core coolant flow rate can be suppressed before the need to scram the reactor arises, and thus operation of the reactor can be continued. The circuit 86 is set beforehand at a power level of coolant block threshold by analysis in such a manner that when the power level is caused to rise by an increase in flow rate, the blocking or running-back can be effected to keep the core characteristics parameters such as the maximum linear heat generating rate and minimum critical power ratio from reaching their critical levels that may cause the breakdown of the fuel cladding. In the embodiment shown in FIG. 2, the coolant block threshold line 88 has the same starting point D as the flow rate control line 38 and is generally situated slightly above line 38. More specifically, the threshold level is about 105% of the rated power level at the rated flow rate and at flow rates adjacent to the rated flow rate and is about 103% of the power level on the flow rate control line 38 in a substantial range of flow rates below these flow rates. Generally, the threshold level at the rated flow rate and flow rates adjacent to the rated flow rate can be set at a value in the range between 103 and 108% of the rated power level, and the threshold level in a substantial range of flow rates below these flow rates can be set at a value in the range between 102 and 107% of the power level for the flow control line 38. As apparent from the foregoing, according to the invention, there is provided, in addition to the APRM 40, TPM 42 and RBM 44 of the conventional core monitoring system, the ORM 46 operative to prevent an excessive rise in the reactor power level caused by an increase in the core coolant flow rate, before the reactor is scrammed. As a result, various advantages are offered in operating a nuclear reactor by the present invention. Firstly, when an operator turns the wrong valves, or some equipment misoperates, for example, the core coolant flow rate may abnormally rise and the power may rapidly rise. When this phenomenon occurs, it is possible to inhibit an abnormal transient change in core characteristics (minimum critical power ratio, maximum linear heat generating rate, rated power, flow vibration characteristics, etc.) by blocking or running back an increase in the core coolant flow rate by controlling the recirculation pumps. Secondly, when TPM 42 and APRM 40 are the only monitoring devices used, the reactor is scrammed when the threshold power level is exceeded as a result of a rise in power caused by an increase in the core coolant flow rate. This makes it inevitable to interrupt the operation of the reactor. However, according to the invention, when the threshold power level (about 105% of rated power) of ORM 46 is exceeded, the increase in the core coolant flow rate is blocked or the flow rate is run-back, so that an excessive rise in power due to an increase in flow rate can be inhibited. After the inhibiting action is performed, the core coolant flow rate can easily be returned to a normal flow rate control condition. Thus the invention minimizes the number of times the reactor is scrammed and enables the reactor to be substantially continuously operated with minimum interruption. Another important advantage offered by the invention is that because of the provision of ORM in addition to TPM and APRM as a system for monitoring the power level caused by a rise of the core coolant flow rate, improvements are provided to the minimum critical power ratio which is the monitor index for preventing the thermal breakdown of the fuel cladding owing to the fact that the scram threshold of TPM and APRM is about 115-120% of the rated power at or near the rated flow rate but ORM has a coolant block and run-back threshold which is about 105% of the rated power and thus the range of variations in minimum critical power ratio before the threshold power level is reached can be reduced to 1/3-1/4 by taking as a reference the range of changes occurring until about 115-120% of the rated power is attained. The same goes for the maximum linear heat generating rate which is the monitor index for preventing the mechanical breakdown of the fuel cladding. Thus as compared with the nuclear reactor having no ORM as disclosed in the aforesaid U.S. Pat. No. 3,565,760, for example, the reactor provided with ORM according to the invention shows no increase in the core characteristics parameters such as minimum critical power ratio and maximum linear heat generating rate above their critical levels which might brought about the breakdown of the fuel cladding, even if the power level is raised in rated operation. Thus a nuclear reactor with ORM could develop higher power than a nuclear reactor of the same design having no ORM. This feature of the invention will be described in detail by referring to the drawings. Generally, in designing a nuclear reactor, the critical lever Lu of any one of core characteristics parameters that may brought about breakdown of the fuel cladding shown in FIG. 7 is first obtained. Then, the operation critical level Lo of the core characteristics parameter for normal operation is set such that critical lever Lu can be maintained even if an excessive rise in power is caused by the carelessness of an operator or misoperation of some equipment. More specifically, the operation critical level Lo is set in such a manner that, assumming that the core characteristics parameter X vary as indicated by a line (a) in FIG. 7 and the range of variations of the core characteristics parameter are denoted by .DELTA.X, then Lo.ltoreq.Lu-.DELTA.X. In FIG. 7, a line (b) represents an unallowable operation condition, and a line (c) is an allowable operation condition in which operation efficiency is lower than in the operation condition represented by line (a). It is essential that in setting the operation critical level Lo, all the factors concerned in a rise in power and all the core characteristics parameters that constitute indices of breakdown of the fuel cladding should be taken into consideration. The principal factors concerned in a rise in power include withdrawing of control rods, an increase in the core coolant flow rate and a rise in the pressure in the core due to shutoff of the load. The indices of breakdown of the fuel cladding include the maximum linear heat generating rate and minimum critical power ratio. The latter can be expressed in terms of the fuel assembly power. FIG. 8 show variations .DELTA.P.sub.L of the maximum linear heat generating rate R.sub.L and variatins .DELTA.P.sub.B of the fuel assembly power P.sub.B occurring in a nuclear reactor provided with APRM and PBM when a rise in power is caused by the three factors referred to hereinabove. FIG. 9 is a view similar to FIG. 8 but showing the values obtained with a nuclear reactor provided with ORM according to the invention in addition to APRM and RBM. As can be clearly seen in FIG. 8, the provision of APRM and RBM enables .DELTA.P.sub.L and .DELTA.P.sub.B to be reduced as indicated by hatching when an excessive power rise is caused by control rod withdrawing and pressure rise, but .DELTA.P.sub.L and .DELTA.P.sub.B show no reduction when an excessive power rise is caused by an increase in the core coolant flow rate. This makes it inevitable to set the operation critical level L.sub.o for normal operation of the reactor by taking into consideration such relatively large values of .DELTA.P.sub.L and .DELTA.P.sub.B. Thus L.sub.o is limited to a low level after all. On the other hand, if ORM is additionally provided .DELTA.P.sub.L and .DELTA.P.sub.B can be reduced in all aspects and thus the operation power level of the reactor can be set at a high level. A further important advantage of the invention is that since the threshold power level at which ORM is set is determined as a function of the core coolant flow rate, a rise in power can be prevented by all means when the power level reaches the threshold power level corresponding to the prevalling flow rate regardless of the situation in which the power is increased by a rise in the core coolant flow rate. This feature of the invention will be described in detail by referring to FIGS. 10 and 11 and by comparing the power monitoring system according to the invention with the control system disclosed in Japanese Patent Publication No. 21518/79 referred to hereinabove in the background of the invention. The control system of the prior art is provided with means for resetting, in a normal operation mode, the recirculation coolant flow rate threshold M and core coolant flow rate threshold C only when the power density calculated at certain time intervals is higher in level than the value obtained by the preceding calculation, to thereby avoid an increase in flow rate above the threshold levels. In this control system, when the power level is reduced by reducing the core coolant flow rate after the threshold levels M and C are set at a high power level P.sub.H shown in FIG. 10 following a slow and gradual rise in power, the threshold level M would be kept at the high level. If, for example, the flow rate rises due to the failure of the flow control system after the period of a low power P.sub.L has lasted for some time, the flow rate would continue to rise until the level M or C is reached. A power level P.sub.H * attained at this time would be higher than the aforesaid high power level P.sub.H by an amount corresponding to a reduction in the amount of Xenon (neutron absorber) in the core occurring during the time the reactor is operated at the low power level P.sub.L. In the case of a reactor provided with ORM according to the invention, when the flow rate begins to rise from the low power P.sub.L under similar circumstances, the power level does not rise above the power level of coolant block threshold as shown in FIG. 11 and the rise in power is blocked at a threshold level P.sub.T corresponding to the prevailing flow rate. That is, according to the invention, even if the power level drops or the amount of Xenon shows a variation prior to the rise in power, it is possible to effectively suppress an excessive rise in power due to a rise in flow rate.
041815726	description	DESCRIPTION OF THE PREFERRED EMBODIMENT In designs of nuclear reactors that utilize rotatable plugs in the closure head of a reactor vessel, there exist annular passageways defined between the rotatable plugs and associated apparatus which allow the rotation of the plugs. In order to meet design requirements, seals must be provided that are capable of sealing these passageways under a variety of reactor conditions. The invention described herein serves to seal those kinds of passageways. Referring to FIG. 1, a core 10 comprising fuel assemblies 12 that produce heat is contained within a reactor vessel 14. The reactor vessel 14 has an inlet 16 and an outlet 18 that permit a coolant 20 to circulate in a heat transfer relationship with the fuel assemblies 12. The coolant 20, which in a fast breeder reactor may be liquid sodium, fills the reactor vessel 14 to a coolant level 22. The reactor vessel 14 is closed at its top end by a closure head comprising a stationary outer ring 24, a first rotatable plug 26, a second rotatable plug 28, and a third rotatable plug 30. The stationary outer ring 24 may be integral with reactor vessel 14 or it may be attached to reactor vessel 14 by suitable means known in the art. In addition, a gear mechanism 32 may be mounted on each rotatable plug such that gear mechanism 32 may be driven by a drive mechanism (not shown) which in turn will rotate the particular rotatable plug. The first rotatable plug 26 is supported from stationary outer ring 24 by a first bearing assembly 34. The outer peripheral surface of the first rotatable plug 26 together with the inner peripheral surface of stationary outer ring 24 define a first annulus 36 therebetween. The first bearing assembly 34 enables the first rotatable plug 26 to move relative to stationary outer ring 24 while maintaining a fluid-tight boundary between the outside and the inside of reactor vessel 14. Again referring to FIG. 1, second rotatable plug 28 is dispsosed eccentrically within first rotatable plug 26 and supported by second bearing assembly 38 defining a second annulus 40 therebetween in a manner similar to that of first bearing assembly 34. Likewise, third rotatable plug 30 is similarly eccentrically disposed within second rotatable plug 28 and supported by a third bearing assembly 42 defining a third annulus 44 therebetween. In addition, third rotatable plug 30 has disposed therein an in-vessel transfer collar 46 which provides access for an in-vessel transfer machine (not shown). During refueling, an in-vessel transfer machine which may be chosen from those wellknown in the art, is placed in the bore of the in-vessel transfer collar 46. When the in-vessel transfer machine is in place in the in-vessel transfer collar 46, a selected combination of rotations of the three rotatable plugs 26, 28, and 30 will align the in-vessel transfer machine in appropriate relationship with a chosen fuel assembly 12 of the core 10. As is well known in the art, the in-vessel transfer machine may then remove the chosen fuel assembly from the core and replace it with a fresh fuel assembly. When the reactor coolant 20 is liquid sodium, as in the case of liquid metal fast breeder reactors, it is necessary to avoid contact of the liquid sodium with oxygen because this interaction will result in the formation of impurities in the liquid sodium. To thus avoid this interaction, the space between the bottom of the closure head and the coolant level 22 is filled with a cover gas 48 such as argon. The cover gas 48 not only fills the cover gas space between the bottom of the closure head and the top of the coolant level 22, but it also fills the annuli 35, 40, and 44. While the cover gas 48 prevents oxygen from contacting the coolant 20, the cover gas 48 itself is subjected to radiation exposure from the core and thus becomes contaminated with radioactive particles. It is, therefore, necessary to have the cover gas 48 circulated between the reactor vessel and the cleaning process to remove most of the radioactive particles in a manner well known in the art. As previously indicated, it is, nevertheless, necessary to prevent this cover gas 48 from escaping up the annuli 36, 40, and 44 through the seals in the closure head, and out of the reactor vessel. Referring now to FIGS. 2 and 3, while FIG. 2 is a partial view of the closure head of FIG. 1 illustrating the three bearing assemblies 34, 38, and 42, FIG. 3 is an enlargement of the first bearing assembly 34 which shows the elements of a typical bearing assembly. The bearing assembly comprises a bearing support 48 which rests upon and is sealed to the stationary outer ring 24 by two O-rings 50 which may be chosen from those well known in the art. The bearing inner race 52 is supported by the bearing support 48 and is bolted thereto by bolt 54. Bearing ball 56 is disposed in inner race 52 in a manner such that an additional clearance 58 is provided on the inner diameter of the inner race 52. Clearance 58 is provided to accommodate differential thermal expansion among the components of the closure head which allows the closure head to be manufactured of materials having various coefficients of thermal expansion. An outer race 60 rests on the bearing ball 56 and is bolted to the first rotatable plug flange 62 by bolt 64. Flange 62 may be attached to the first rotatable plug 26 by common means such as bolts or flange 62 may be an integral part of first rotatable plug 26. The arrangement of the first bearing assembly 34 and flange 62 of first rotatable plug 26 is such that the weight of first rotatable plug 26 is transmitted through the flange 62 and through bearing assembly 34 to the stationary outer ring 24 thereby providing a mechanism for allowing rotation of the first rotatable plug 26 with respect to the stationary outer ring 24 along annulus 36. A spacer 66 attached to inner race 52 is provided to maintain proper alignment of bearing ball 56. The configuration of bearing support 48 in conjunction with the configuration of stationary outer ring 24 define two reservoirs between them; a first reservoir 68 and a second reservoir 70. In addition, the configuration of the components of bearing assembly 34 further define first annulus 36. Still referring to FIG. 3, an outer seal 72 is disposed on the bearing support 48 so as to seal the annulus between flange 62 and bearing support 48. Outer seal 72 comprises a tubular seal element 74 which may be a stainless steel hollow O-ring disposed in annulus 36 that extends the circumference of the stationary outer ring 24 and load spring assembly 76 attached to the bearing support 48 so as to force the tubular seal element 74 against flange 62 thereby sealing the annulus. Outer seal 72 may be disposed on the bearing support 48 in various configurations; however, the preferred angle is approximately 30 degrees from the vertical. Likewise, an inner seal 78 is similarly disposed on the bearing support 48. Inner seal 78 also comprises a second tubular seal element 80 which may also be a stainless steel O-ring and a second load spring assembly 82 attached to the bearing support 48 so as to compress the second tubular seal element 80 against flange 62. A lubricant inlet 84 which may be a conduit chosen from those well known in the art is disposed in stationary outer ring 24 and bearing support 48 such that the outlet of lubricant inlet 84 is disposed on the underside of inner race 52. Lubricant inlet 84 is connected on its outer end to a lubricant pump 86 which may be a constant volume pump which is capable of pumping a lubricant such as silicone through the lubricant inlet 84, through a channel 88 where the lubricant flow divides into two flow paths, one flowing through first annulus 36 toward outer seal 72 and the other flowing through first annulus 36 toward bearing ball 56. The lubricating fluid flowing through the second path under pressure is forced around bearing ball 56 and over tubular seal element 80 thereby compressing load springs 82 and allowing the lubricant to pass between the tubular seal element 80, and flange 62. From inner seal 78, the lubricant flows into first reservoir 68 where it fills first reservoir 68 to a level 90. At the same time, the lubricating fluid flows through the first path over outer tubular seal element 74 and into second reservoir 70. As the lubricant passes over the tubular seal elements 74 and 80 a film of lubricant is established between the tubular seal element and flange 62 such that no gases may pass therebetween. In addition, the force of the lubricant on flange 62 can reduce the bearing load by as much as 10 to 20%. A typical silicone lubricant may be Dow Corning No. 710 cracked at 482.degree. F. to remove low volatility fractions. The cracking avoids most of the off gassing at 450.degree. F., the seal operating temperature. Still referring to FIG. 3, a return conduit 92 is connected between second reservoir 70 and a first valve 94 which may be a three-way valve chosen from those well known in the art while another return conduit 96 is provided between first reservoir 68 and first valve 94. The return conduits 92 and 96 serve to direct the lubricating fluid to first valve 94 where the lubricating fluid is recirculated to lubricant pump 86. Furthermore, a recirculating conduit 98 is connected to lubricant inlet 84 and around lubricant pump 86 with a gate valve 100 and a pressure relief valve 102 disposed therein to enable lubricant pump 86 to maintain a constant volume flow even under varying operating conditions. However, during reactor refueling gate valve 100 is closed which prevents flow in recirculating conduit 98 and results in increased pressure on flange 62 which reduces the load on the bearing at a time when it is necessary to rotate the plugs. In addition, a gas inlet line 104 is connected to first annulus 36 while a gas outlet line 106 is disposed in first reservoir 68 with an opening above lubricant level 90 so that a gas such as argon may be pumped through first annulus 36 to thereby entrain contaminants in the gas flow thus purging the annulus. Also a check valve 108 may be disposed in gas inlet line 104 to prevent reverse flow in that line. Still referring to FIG. 3, it should be noted that a circumferential extension 110 of flange 62 extends into bearing support 48 thereby defining a liquid dip seal 112 in first annulus 36. While the lubricating fluid is being pumped through first annulus 36 the lubricating fluid fills liquid dip seal 112 creating a fluid seal against gas leakage through first annulus 36. Moreover, should lubricant pump 86 not be operating, the lubricating fluid will, nevertheless, remain in liquid dip seal 112 thus sealing the annulus even when the lubricating fluid is not flowing. Referring now to FIG. 4, a typical load spring assembly such as load spring assembly 82 comprises a housing 114, a biasing mechanism such as a coil spring 116 mounted in housing 114, a platform 118 mounted in housing 114 on an end of coil spring 116, and a contact surface 120 attached to platform 118 for contacting tubular seal element 80. Coil spring 116 serves to force the tubular seal element against a surface such as flange 62 to seal the annulus 36. Of course, under pressure from the lubricating fluid coil spring 116 may be compressed thereby relieving pressure on the tubular seal element. Therefore, the invention provides a closure head for a nuclear reactor having a sealing and lubricating system for allowing rotation of rotatable closure head plus while sealing the annuli defined by the rotatable plugs.
055486258	claims	1. A method for performing parallel multiple field processing in x-ray lithography comprising the steps of: a) placing a semiconductor wafer on a support stage for holding the semiconductor wafer; b) providing an x-ray source; c) providing a mirror assembly containing at least two mirrored surfaces each having a length; d) providing means for containing the mirror assembly, wherein the length of the at least two mirrored surfaces is positioned along a length of the means for containing the mirror assembly to act as collimating mirrors; e) providing a mask assembly containing a same number of masks as mirrored surfaces in the mirror assembly, wherein the mask assembly maintains a first fixed separation distance between each mask for forming multiple and separate image fields; f) aligning the semiconductor wafer with respect to the mask assembly; g) focusing a first surface portion of the semiconductor wafer with respect to the mask assembly; h) exposing the first surface portion of the semiconductor wafer to at least two exiting x-ray beams formed by the collimating mirrors which collimate photons emitted by the x-ray source which is projected into a first end of the means for containing the mirror assembly such that the at least two exiting x-ray beams are collimated and reflected off the at least two mirrored surfaces to travel through the length of the means for containing the mirror assembly to exit a second end of the means for containing the mirror assembly, wherein the at least two mirrored surfaces form the at least two exiting x-ray beams having a second fixed separation distance between the at least two exiting x-ray beams such that each of the at least two exiting x-ray beams travels through a separate mask of the mask assembly to print the multiple and separate image fields on the semiconductor wafer; and i) stepping to a next surface portion of the semiconductor wafer and repeating steps (g) through (i) until all desired surface portions of the semiconductor wafer are exposed. a) placing a semiconductor wafer on a support stage for holding the semiconductor wafer; b) providing an x-ray source; c) providing a monolithic mirror assembly containing at least two mirrored surfaces each having a length; d) providing an elongated tube, suitably evacuated, for containing the monolithic mirror assembly, wherein the length of the at least two mirrored surfaces is positioned along a length of the elongated tube to act as collimating mirrors; e) providing, external to the elongated tube but aligned thereto, a mask assembly containing a same number of masks as mirrored surfaces in the monolithic mirror assembly, wherein the mask assembly maintains a first fixed separation distance between each mask for forming multiple and separate image fields; f) aligning the semiconductor wafer with respect to the mask assembly; g) focusing a first surface portion of the semiconductor wafer with respect to the mask assembly; h) exposing the first surface portion of the semiconductor wafer to at least two exiting x-ray beams formed by the collimating mirrors which collimate photons emitted by the x-ray source which is projected into a first end of the elongated tube for containing the monolithic mirror assembly such that the at least two exiting x-ray beams are collimated and reflected off the at least two mirrored surfaces to travel through the length of the elongated tube to exit a second end of the elongated tube for containing the monolithic mirror assembly, wherein the at least two mirrored surfaces form the at least two exiting x-ray beams having a second fixed separation distance between the at least two exiting x-ray beams such that each of the at least two exiting x-ray beams travels through a separate mask of the mask assembly to print the multiple and separate image fields on the semiconductor wafer; and i) stepping to a next surface portion of the semiconductor wafer and repeating steps (g) through (i) until all desired surface portions of the semiconductor wafer are exposed. a) placing a semiconductor wafer on a support stage for holding the semiconductor wafer; b) providing an x-ray source; c) providing a coupled mirror assembly containing at least two mirrors each having a length; d) providing an elongated tube, suitably evacuated, for containing the coupled mirror assembly, wherein the length of the at least two mirrors is positioned along a length of the elongated tube to act as collimating mirrors; e) providing, external to the elongated tube but aligned thereto, a mask assembly containing a same number of masks as mirrors in the coupled mirror assembly, wherein the mask assembly maintains a first fixed separation distance between each mask for forming multiple and separate image fields; f) aligning the semiconductor wafer with respect to the mask assembly; g) focusing a first surface portion of the semiconductor wafer with respect to the mask assembly; h) exposing the first surface portion of the semiconductor wafer to at least two exiting x-ray beams formed by the collimating mirrors which collimate photons emitted by the x-ray source which is projected into a first end of the elongated tube for containing the coupled mirror assembly such that the at least two exiting x-ray beams are collimated and reflected off the at least two mirrors to travel through the length of the elongated tube to exit a second end of the elongated tube for containing the coupled mirror assembly, wherein the at least two mirrors form the at least two exiting x-ray beams having a second fixed separation distance between the at least two exiting x-ray beams such that each of the at least two exiting x-ray beams travels through a separate mask of the mask assembly to print the multiple and separate image fields on the semiconductor wafer; and i) stepping to a next surface portion of the semiconductor wafer and repeating steps (g) through (i) until all desired surface portions of the semiconductor wafer are exposed. 2. The method of claim 1, wherein the step of placing the semiconductor wafer is further characterized as placing a wafer having a diameter of at least 200 millimeters. 3. The method of claim 1, wherein the step of providing the means for containing the mirror assembly provides an evacuated tube having a multiple beryllium exit window at the second end of the means for containing the mirror assembly. 4. The method of claim 1, wherein the step of providing the mirror assembly comprises providing a monolithic mirror having multiple mirrored surfaces having a shape selected from a group consisting of: flat, parabolic, circular, toroidal, cylindrical, and polynomial. 5. The method of claim 1, wherein the step of providing the mirror assembly comprises providing multiple separate mirrors mechanically coupled together and having a fixed physical separation between each mirror, wherein each mirror has a surface having a shape selected from a group consisting of: flat, parabolic, circular, toroidal, cylindrical, and polynomial. 6. The method of claim 1, wherein the step of providing the mask assembly provides multiple masks coupled together with means for coupling selected from a group consisting of: an interferometer, physical means, and electronic means. 7. The method of claim 1, wherein the step of providing the mask assembly provides a first mask used for aligning and focusing, and a second mask used only for focusing. 8. The method of claim 1, wherein the step of providing the mask assembly provides masks that are each used for aligning and focusing. 9. A method for performing parallel multiple field processing in x-ray lithography comprising the steps of: 10. The method of claim 9, wherein the step of placing the semiconductor wafer is further characterized as placing a wafer having a diameter of at least 200 millimeters. 11. The method of claim 9, wherein the step of providing the mask assembly provides multiple masks coupled together with means for coupling selected from a group consisting of: an interferometer, physical means, and electronic means. 12. The method of claim 9, wherein the step of providing the monolithic mirror assembly provides mirrored surfaces having a shape selected from a group consisting of: flat, parabolic, circular, toroidal, cylindrical, and polynomial. 13. The method of claim 9, wherein the step of providing the elongated tube for containing the monolithic mirror assembly comprises providing an ultra high vacuum tube, evacuated to a pressure of approximately in a range of 10.sup.-10 to 10.sup.-11 torr. 14. The method of claim 13, wherein the step of providing the elongated tube further provides a multiple beryllium exit window at the second end of the elongated tube. 15. The method of claim 9, wherein the step of providing the mask assembly provides a first mask used for aligning and focusing, and a second mask used only for focusing. 16. A method for performing parallel multiple field processing in x-ray lithography comprising the steps of: 17. The method of claim 16, wherein the step of placing the semiconductor wafer is further characterized as placing a wafer having a diameter of at least 200 millimeters. 18. The method of claim 16, wherein the step of providing the coupled mirror assembly comprises providing multiple separate mirrors mechanically coupled together to maintain a fixed physical separation between each mirror. 19. The method of claim 16, wherein the step of providing the coupled mirror assembly provides mirrors having a shape selected from a group consisting of: flat, parabolic, circular, toroidal, cylindrical, and polynomial. 20. The method of claim 16, wherein the step of providing a mask assembly provides multiple masks coupled together with means for coupling selected from a group consisting of: an interferometer, physical means, and electronic means.
	abstract	An X-ray generator for generating plasma and X-ray emitted from the plasma includes a unit for generating the plasma, and plural reflection optical systems for introducing the X-ray through different optical paths.
	abstract	A heat removal system for the under reactor pressure vessel area of a boiling water nuclear reactor system that provides both protection of the containment boundary from attack by molten core debris and cools the molten core debris to prevent a breach of the containment boundary in the unlikely event of a severe accident where the molten core penetrates the lower head of the reactor vessel is described. The heat removal system includes a glass matrix slab positioned adjacent the floor of the containment and a plurality of heat tubes at least partially embedded in the glass matrix slab and extending into the area under the nuclear reactor pressure vessel. The cooling system also includes a passive containment cooling system, and fused vent pipes connecting the suppression pool with the drywell.
061513768	abstract	The present invention relates to a fuel assembly with a substantially square cross section for a light-water reactor. The light-water reactor comprises a plurality of fuel rods extending between a top tie plate and a bottom tie plate. A fuel rod comprises a cladding tube which surrounds a column with fissionable material. According to one aspect of the invention, at least one fuel rod is provided with an axial gap in the fissionable material. Around this axial gap, the mass of the material is greater than the mass of the material in the main part of the cladding tube.
	description	This is a continuation, under 35 U.S.C. §120, of copending international application PCT/EP2004/006837, filed Jun. 24, 2004, which designated the United States; this application also claims the priority, under 35 U.S.C. §119, of German patent application No. 103 28 773.6, filed Jun. 25, 2003; the prior applications are herewith incorporated by reference in their entirety. The invention lies in the field of nuclear engineering. More specifically, the invention relates to a nuclear system with a safety containment to which a pressure relief line is connected. It relates, further, to a method for the decompression of a system of this type. In a nuclear power plants, incidents or accident situations must be expected to entail a possibly significant pressure rise within the safety containment, depending on the respective incident and on countermeasures initiated where appropriate, such as, for example, the inertization of the containment atmosphere. In order to avoid possibly resulting structural damage to the safety containment per se or even to system components arranged in it, nuclear power stations can be designed for an on-demand depressurization of the containment by the discharge of containment atmosphere (venting). For this purpose, conventionally, a pressure relief line is connected to the safety containment of a nuclear system. The containment atmosphere, however, normally contains radioactive material, such as, for example, noble gases, iodine or aerosol, which could pass into the surroundings of the nuclear power station during venting. Particularly in the case of comparatively serious incidents with the possible occurrence of core melt, airborne activity quantities (aerosols) may arise in particularly high concentrations within the containment, so that, if there are serious leaks or if inadmissible overpressure situations arise, a release of significant quantities of such aerosols or activity quantities into the surroundings of the nuclear system could occur. Airborne activities of this type could cause comparatively long-term land contamination, particularly on account of the high half-lives of possibly entrained components, such as, for example, iodine or cesium isotopes. In order to avoid this, the depressurization systems provided for venting the containment atmosphere are conventionally provided with filter or retention devices which are intended to prevent a release of airborne activity quantities entrained in the containment atmosphere into the surroundings. For this purpose, for example, European patent EP 0 285 845 B1 and U.S. Pat. No. 4,873,050 disclose a concept for the pressure relief of a nuclear power station, in which a Venturi scrubber provided as a filter for the retention of airborne activities and also a throttle device are connected in series into a pressure relief line connected to the safety containment of the nuclear power station. The Venturi scrubber in this case comprises a number of Venturi tubes which are disposed in a washing liquid reserved in a vessel and which can be acted upon by the gas stream carried in the pressure relief line. The Venturi tubes in this case each comprise a nozzle-like constriction at which the gas stream flowing through is accelerated to a particularly high flow velocity. In the region of this constriction point, ingress ports are provided for the washing liquid, the entering washing liquid being entrained by the gas stream flowing through. Owing to the comparatively high flow velocity of the gas stream at this point, a fragmentation of the washing liquid takes place, airborne activities or aerosols entrained in the gas stream being introduced into the liquid droplets thereby occurring. Thus, as a result of a subsequent droplet separation from the gas stream, it is possible to remove a large part of the entrained aerosols or airborne activities. In the system described in EP 0 285 845 B1 and U.S. Pat. No. 4,873,050, the throttle device connected in series with the Venturi scrubber is designed for operation with what is known as critical depressurization. In critical depressurization, the pressure conditions in the line system, that is to say, in particular, the pressure drop across the throttle device, are established such that the medium flowing in the line flows through the throttle device at sound velocity. In the system according to EP 0 285 845 B1 and U.S. Pat. No. 4,873,050, this effect is utilized, in a response situation, that is to say during a depressurization of the containment, to set a volumetric throughput in the pressure relief line which is constant over time. It is accordingly an object of the invention to provide a nuclear installation and a pressure relief method for a nuclear installation which overcomes the above-mentioned disadvantages of the heretofore-known devices and methods of this general type and, in which, in the event of depressurization, even the finest possible airborne activities or aerosols are retained in the Venturi scrubber with particularly high reliability, so that a release into the surroundings is ruled out with particularly high reliability. Furthermore, a method for the depressurization of a nuclear system of this type is to be specified. With the foregoing and other objects in view there is provided, in accordance with the invention, a nuclear plant, comprising: a containment; a pressure relief line communicating with said containment; a throttle device and a Venturi scrubber connected in series in said pressure relief line, said Venturi scrubber being disposed in a vessel with a washing liquid; said Venturi scrubber and said throttle device being dimensioned to establish, in the event of a critical depressurization of an air/vapor mixture flowing in said pressure relief line and through said throttle device, a flow velocity of the air/vapor mixture of more than 150 m/s, or above 200 m/s, in the Venturi scrubber. In other words, the objects of the invention are achieved, in that the Venturi scrubber and the throttle device are dimensioned such that, in the event of a critical depressurization of an air/vapor mixture flowing in the pressure relief line, a flow velocity of the air/vapor mixture of more than 150 m/s, preferably of more than 200 m/s, is set at the throttle device in the Venturi scrubber. Dimensioning in this case preferably takes place in such a way that this high velocity prevails predominantly in the entire operating overpressure range of the separation device, independently of the respective operating pressure of, for example, 2-10 bar. The Venturi pressure losses of, for example, >0.5 bar at 1 bar and, for example, >2 bar at pressures >5 bar, which occur in the case of a higher operating pressure of the washer device for generating the corresponding acceleration of the gases of higher density, are in this case set passively over the entire operating range as a result of the combination of the Venturi scrubber and the throttle. The invention proceeds in this case from the consideration that, for the separation of airborne activities or aerosols, a comparatively fine droplet mist is generated in a Venturi scrubber or a Venturi tube as a result of the flow conditions prevailing in the tube interior when water is fed into the latter, the airborne activities or aerosols to be separated being capable of being introduced into such droplets and therefore of being removed from the gas stream together with these. A particularly high separation action even for the finest possible aerosols can thus be achieved, in that the probability with which the aerosols impinge onto suitable water droplets, in particular with the assistance of correspondingly high washing liquid loads, and are included in these water droplets is kept particularly high. As surprisingly became apparent, precisely with regard to Venturi tubes in which the feed of the washing liquid into the tube interior is ensured in the manner of a passive type of construction via the underpressure prevailing at the constriction point and therefore without external drive means, the probability of impingement and inclusion of even the finest possible aerosols in the droplet mist rises to a considerable, highly superproportional extent, so that, in the case of very high flow velocities of the gas stream in the Venturi tube, separation rates for mixed aerosols with a particle size of about 1 μm of more than 99.9% and for comparatively fine aerosols with a particle size of less than 0.5 μm of 98% and above can be achieved in the washing liquid. The depressurization and activity retention system of the nuclear system is therefore designed for maintaining such high flow velocities in the depressurization situation. In order in this case, precisely with regard to the characteristic parameters, such as, for example, system pressure, which possibly change to a great extent in the event of an incident scenario over the entire course of the incident, to ensure such a high separation rate in every phase of a possible incident and therefore to prevent to the greatest possible extent a release of contaminating constituents into the surroundings in every phase of an incident, the depressurization and activity retention system of the nuclear system is moreover designed for such a high degree of separation virtually independently of the system pressure prevailing in the safety containment of the nuclear system. In this case, deliberate use is made of the knowledge that, in the case of a throttle device operating with what is known as critical depressurization, the flow medium flows through said throttle device at its sound velocity independently of the prevailing inlet pressure. Thus, in the state of critical depressurization, the volumetric throughput through the throttle device is constant independently of the prevailing inlet pressure. A suitable combination of the Venturi scrubber with the throttle device and, if appropriate, with a metallic fine aerosol follow-up filter can thus ensure that, in the event of critical depressurization via the throttle device, the volumetric throughput of the flowing medium through the Venturi scrubber and, if appropriate, through the fine aerosol follow-up filter can be kept virtually constant independently of the system pressure prevailing in the safety containment and transferred to the inlet side of the throttle device. Thus, by the throttle device being combined with the Venturi scrubber, a uniformly high degree of separation at the Venturi scrubber and, if appropriate, at the fine aerosol follow-up filter can be ensured virtually over the entire incident scenario, to be precise as long as critical depressurization via the throttle device occurs due to the prevailing pressure conditions. For this purpose, the Venturi scrubber and the throttle device are in each case suitably dimensioned in the manner of coordination with one another, so that, in the case of critical depressurization occurring at the throttle device, the desired flow conditions with a particularly high flow velocity in the Venturi scrubber and, if appropriate, an optimum velocity in the fine aerosol follow-up filter are established. The minimum flow velocity of the flow medium in the Venturi scrubber which is required for the desired high degree of separation may in this case depend on the exact composition of the flow medium and may shift toward higher values in the case of changing gas compositions, for example in the case of a higher H2 fraction. As became apparent, however, a sufficiently high degree of separation can be achieved for the flow media possibly occurring in the event of the depressurization of the safety containment of a nuclear system, in that the combination of the Venturi scrubber and the throttle device is designed and dimensioned in the manner of a calibration or reference, in such a way that, in the case of an air/vapor mixture flowing in the pressure relief line, with critical depressurization occurring at the throttle device, a flow velocity of the air/vapor mixture of more than 150 m/s, preferably of more than 200 m/s, prevails in the Venturi scrubber. The flow velocity of the flow medium is in this case determined particularly in the region of the constriction point of the respective Venturi tube. The high velocity set via the combination of the Venturi scrubber and throttle device may shift toward even higher values in the event of changing gas compositions, for example a higher H2 fraction, on account of the higher sound velocity. It was found, furthermore, that a critical maximum velocity of approximately 270-300 m/s is established in the two-phase mixture consisting of gas mixture and washing liquid in the Venturi scrubber. Owing to the preferred selection of a particularly high Venturi design velocity of, for example, 200 m/s, which corresponds approximately to ⅔ of the maximum two-phase mixture velocity of about 300 m/s, it can thus be ensured that, even in the presence of mixtures with a higher sound velocity, inherently reliable throughput limitation becomes possible and the following retention devices are reliably protected against overload. Advantageously, the Venturi scrubber comprises a plurality of Venturi tubes. These may be designed as what are known as short Venturi tubes, the outlets of which are arranged below the intended desired level of the washing liquid, so that the Venturi tubes are immersed essentially completely in the washing liquid. In this case, it proves to be particularly beneficial that the higher-lying separator filter section is protected by an overflow weir against the water backwash occurring, so that, in this variant, too, a reduced component height becomes possible. In this embodiment, a combination with a following metal fiber filter proves to be particularly advantageous for a particularly high overall separation. By means of Venturi nozzle tubes ejecting primarily above the washing liquid, the water backwash determining the component size can be minimized, and, furthermore, a markedly higher empty tube velocity can be set in the Venturi washing device. This results in a considerably smaller Venturi scrubber diameter and a smaller component height and also a correspondingly reduced consumption of washing liquid. Due to the compact type of construction made possible thereby, particularly in combination with existing water reservoirs, the easy integration of the device even in particularly protected building parts of the system, such as, for example, the reactor building, along with a reduced outlay in terms of shielding, becomes possible. Advantageously, a comparatively large fraction of the Venturi tubes is therefore designed as what are known as long Venturi tubes, the outlets of which are arranged above the intended desired level of the washing liquid. In order, furthermore, to prevent a sedimentation in the region of the vessel, which could lead to an increased maintenance and care requirement, in a further advantageous embodiment the Venturi scrubber is designed for a comparatively intensive swirling and circulation of the washing liquid in the operating situation. For this purpose, advantageously, a small fraction of the Venturi tubes, preferably up to about 10%, is arranged with a downwardly directed outlet direction within the vessel and below the desired level of the washing liquid. It has proved particularly beneficial for ensuring high separation rates to set a comparatively high water load in the Venturi scrubber of, for example, more than 5 liters, preferably more than 10 liters, of washing liquid per cubic meter of gas. In order to ensure this, in a further advantageous embodiment, the Venturi tubes have an annular slit feed extending over the nozzle circumference and having an opening angle of 20° to 85°, preferably of 30° to 45°. For such a high water load, furthermore, the Venturi tubes of the Venturi scrubber advantageously have in each case a ratio of their neck cross-sectional area to the inlet area for the washing liquid of less than 10:1, preferably of about 3:1. The neck cross-sectional area in this case indicates the cross-sectional area, through which the flow medium can flow freely, at the constriction point within the respective Venturi tube. In a particularly advantageous embodiment, the Venturi tubes of the Venturi scrubber are designed in such a way that the passive intake and distribution of washing liquid are ensured into the core jet region inside the Venturi tube on account of the underpressure generated by the medium flowing through. For this purpose, the Venturi tubes of the Venturi scrubber are advantageously designed as round Venturi tubes with a neck width of less than about 80 mm, preferably of less than about 40 mm, or as flat Venturi nozzles with a neck width of less than about 100 mm. Additionally, or alternatively, the Venturi tubes of the Venturi scrubber advantageously have a ratio of height to neck width of more than 5, preferably of more than 10. A particularly compact type of construction for the depressurization and activity retention system assigned to the nuclear system, with a correspondingly reduced outlay in terms of production and of assembly and with the capability of easy accommodation in the protected system region, can be achieved in that the vessel equipped with the Venturi scrubber is advantageously connected on the washing-fluid side to a further washing liquid store. Consequently, the quantity of washing liquid reserved in the vessel itself can be kept comparatively small, while, if required, that is to say, in particular, in the case of the occurrence of a consumption of washing liquid, an on-demand afterfeed from the further washing liquid store may be provided. The in this sense inactive, in particular larger, washing liquid reservoir may in this case be stored in a separate storage vessel and, in particular, serve for topping up evaporated washing liquid. The filling level in the vessel may in this case be set passively by the further washing liquid store being arranged at the same geodetic height or by means of a filling-level float control. In this case, in particular, further water reservoirs already provided in any case, such as, for example, wastewater tanks, demineralized water supply or the like, may also be utilized as a further washing liquid store, while the on-demand feed of washing liquid into the vessel may take place via gradients or by means of diaphragm pumps operated from a compressed-air accumulator, independently of the possibly failed power supply. Particularly effective activity retention can be achieved in that the depressurization and activity retention system assigned to the nuclear system is designed, in a particularly advantageous embodiment, for an on-demand recirculation of the airborne activities or aerosols separated in the washing liquid into the containment. For this purpose, in a particularly advantageous embodiment, the vessel provided with the Venturi scrubber is connected on the washing-fluid side to the interior of the safety containment via a feedback line of the nuclear system. By virtue of such an embodiment, if required, that is to say, in particular, constantly or at cyclic intervals, the washing liquid laden with activities or aerosols removed from the gas stream can be displaced completely or partially into the safety containment, so that the activity overall requiring treatment remains reliably in the containment. By virtue of the activity reduction in the washing liquid achieved by means of the feedback, resuspension effects occurring, which could lead to the discharge of activity into the following filter devices, are minimized. In this case, an afterfeed of the washing liquid into the vessel, in particular from the further washing liquid reservoir can take place. As a result of such a recirculation or feedback of the activities, the activity quantity and concentration contained overall in the washing liquid can be kept particularly low, so that, for example, even resuspension effects leading to the discharge of activity into following filter devices can be kept particularly low. As a result, in combination with the high Venturi degree of separation, a reduction in the filter load and consequently in the filter surfaces required is possible. Furthermore, particularly in the case of a comparatively lengthy venting operation over several days, a significant improvement in activity retention, particularly with regard to iodine and with aerosols, can be achieved. Furthermore, as a result of such a feedback or recirculation of the activities separated in the Venturi scrubber, the decay heat occurring via the aerosols or airborne activities is kept away from the vessel and is displaced back into the containment, so that the possible loads occurring as a result of this in the vessel, for example due to fluid evaporation, can be kept particularly low, so that a comparatively lengthy venting operation over several days and weeks thereby becomes possible, without the following metal fiber fine filters being overloaded by resuspension aerosols and iodine separation at the iodine sorption filter being overloaded by iodine resuspension. Such design requirements with long-term venting operation can therefore be fulfilled reliably and particularly cost-effectively even for recent reactor systems with increased requirements with regard to the control of serious incidents, for example in combination with independent washing liquid afterfeed, for example via diaphragm pumps, etc., and activity recirculation into the containment via quantity-limiting throttles. Precisely because the evaporation of washing liquid can thereby be avoided, the overall result, that is to say also taking into account the possibly provided afterfeed of washing liquid into the vessel, is an overall reduced requirement of washing liquid. In order to keep particularly low the number of required leadthroughs through the safety containment of the nuclear system which are designed with a view to considerable safety requirements, in a further advantageous embodiment the feedback line is in this case connected to the interior of the safety containment via the pressure relief line. Recirculation or feedback in this case takes place by jet feed into the central region of the pressure relief line, so that a transfer of the activity-laden washing liquid into the containment can take place in countercurrent to the depressurization gas stream. Advantageously, the Venturi section is followed by double gravity-type drop separation with drop recirculation. For drop separation, preferably a centrifugal separator, operated at high speeds >10 m/s, is used, which may at the same time be employed for superheating via the generation of a throttle effect. In the event that there is a following metal filter stage, there is therefore no occurrence of drops, so that this unit may also be arranged, lower-lying or at the same height, thus reducing the space requirement and space height. For further dehumidification and prefiltering, a fiber separator in the exhaust air stream with fibers <50 μm is advantageously combined with a prefilter unit with fibers <20 μm, preferably in decreasing fiber thicknesses. Fine filtering preferably takes place with fibers of up to <5 μm, so that even the small quantity of penetrating fine aerosols of <0.5 μm can still largely be retained. The filter elements are preferably produced from high-grade steel fibers. Fine filtering may also take place with sintered fiber filters having pore diameters <2 μm. For effective organoiodine separation, preferably, a molecular sieve, for example coated with silver nitrate or other silver compounds, etc., is provided, downstream of the throttling, for the long-term operation of the retention system. The superheating of the gas stream before entry into the molecular sieve in this case expediently takes place primarily by throttling, by an amount of at least 50% of the still available pressure gradient of, for example, >2 bar with respect to the maximum operating pressure. Passive and simple superheating of the gas stream in the iodine sorption filter thereby becomes possible. The retention devices, that is to say the Venturi scrubber and the metal fiber filter, may also be accommodated within a vessel at a staggered height, the high-lying filters being provided with an inflow weir, so that a particularly small overall height is obtained. In order in this case to allow feedback in the manner of a completely passive system without recourse to external active components, in a further advantageous embodiment the vessel is arranged so as to lie geodetically at least about 5 m, preferably at least 10 m, higher than the outlet point of the pressure relief line from the safety containment. Consequently, the feedback of the activity-laden washing liquid through the pressure relief line into the containment is possible solely on account of the geodetic pressure in the water column between the pressure relief line and the vessel, so that jet feedback can take place in countercurrent to the gas stream without further active aids. Advantageously, the washing liquid is designed to a particular extent for an effective retention of iodine or iodine-containing compounds. For this purpose, advantageously, a washing liquid with a pH value of at least 9 is reserved in the vessel, and this pH value can be obtained, for example, by the addition of NaOH, other lyes and/or sodium thiosulfate. Adding these chemicals to the washing liquid may advantageously lead to the setting of a concentration in the washing liquid of 0.5 to 5 percent by weight due to intake from a separate chemical vessel via a jet pump located in the freshwater stream. A particularly compact type of construction can be achieved in that, in a further advantageous embodiment, the throttle device is integrated into the vessel. By an additionally provided direct feed of cold water completely or partially via the retention device into the region of the reactor pressure vessel in counter-current to the vent gas, preferably as simple emergency measures by means of existing systems, such as, for example, by means of a firefighting pump, or via other systems, activity recirculation and a cooling of the reactor core by the absorption of energy can advantageously be achieved at the same time. Moreover, since higher feed quantities, with a rising filling level in a containment, occur particularly in the early phase of an accident, a further advantageous reduction in the vapor/gas mixture to be sucked away and therefore, at the same time, a reduction in the dimensions of the retention device or suckaway device can be achieved. As regards the method for the depressurization of a nuclear system of the type mentioned, the object is achieved in that the Venturi scrubber is acted upon by a flow velocity of the medium carried in the pressure relief line of more than 150 m/s, preferably of more than 200 m/s. The advantages achieved by means of the invention are, in particular, that, owing to the deliberate combination of the throttle device with the Venturi scrubber, the mutually coordinated dimensioning essentially over the entire course of an incident can ensure that a particularly high flow velocity of the depressurization gas stream flows through the Venturi scrubber. As a result, in any event, a particularly high separation action of more than 98% of the entrained airborne activities or aerosols already in the washing liquid, in particular even of the fine aerosols with a particle size of less than 0.5 μm, is ensured, so that a release of activities into the surroundings is avoided particularly reliably. The depressurization and activity retention system formed by the Venturi scrubber, the following throttle device and, if appropriate, the metal fiber fine filter in this case, in the manner of a passively operating system, automatically ensures, in virtually all the phases of an incident, an essentially constant through-flow through the Venturi scrubber, independently of the system pressure prevailing in the safety containment, so that this system is suitable, in particular, for what is known as sliding-pressure operation, that is to say for direct action by the system pressure in the safety containment without a further preceding throttle device. Depending on the flow medium carried in the pressure relief line, the virtually constant throughput through the Venturi scrubber can in this case be ensured by the critical depressurization via the throttle device, as a result of which, independently of the prevailing system pressure, the flow velocity of the medium in the throttle device amounts approximately to its sound velocity, so that the volumetric throughput through the Venturi scrubber is correspondingly also constant approximately independently of pressure. In the event of a gas mixture carried in the pressure relief line, moreover, with comparatively high nozzle velocities of 150 m/s to 200 m/s, limited to <300 m/s in the case of, for example, a high H2 fraction, being maintained, a mixture-independent passive throughput limitation through the Venturi scrubber can be ensured even by means of the pressure loss generated by the latter. As a result of the combination of the high-velocity Venturi scrubber device with recirculation, combined with the following metal fiber filters, an overall degree of separation of >99.99 to 99.999% can be ensured even in long-term operation, independently of the aerosol concentration in the containment. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a nuclear system and method for the decompression of a nuclear system, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, there is shown a nuclear plant 1 (also: nuclear engineering installation, nuclear technology plant) with a containment 2 that contains the nuclear components provided for electricity generation and further system components. So that structural damage to or instabilities of the safety containment 2 can be reliably ruled out even in the event of a comparatively serious incident in which a sharp pressure rise within the safety containment 2 must be expected as a result of processes taking place within the containment 2, the nuclear plant 1 is equipped with a depressurization and activity retention system 4 connected to the containment 2. This system makes it possible, as required, to have a deliberate and controlled discharge of containment atmosphere, also designated as venting, out of the safety containment 2 into its surroundings. The depressurization and activity retention system 4 comprises a pressure relief line 6 which is connected to the safety containment 2 and which is connected on the outlet side to a blow-off vent or stack 8. To avoid a contamination of the surroundings of the nuclear plant 1 in the event of a venting or discharge of containment atmosphere, the depressurization and activity retention system 4 is designed for a reliable retention also of airborne activities or aerosols contained in the containment atmosphere. For this purpose, the depressurization and activity retention system 4 comprises a wet scrubber 10 provided as a filter device for such airborne activities or aerosols. The wet scrubber 10 itself comprises a Venturi scrubber 12 which is connected into the pressure relief line 6 and which is disposed in a container or vessel 14 having a washing liquid W. The Venturi scrubber 12 comprises a number of Venturi tubes 16 which issue with their outlets 18 into a gas space 22 located in the vessel 14 above the setpoint level 20 of the washing liquid W. A throttle device 24 is arranged in the gas space 22, and is thus integrated into the vessel 14, and is therefore connected in series with the Venturi scrubber 12 on the gas-stream side. The throttle device 24 is itself connected on the outlet side to a further portion of the pressure relief line 6, said further portion being connected to the blow-off chimney 8 via a filter device 26. The filter device 26 itself comprises a metal fiber filter 28, an intermediate throttle 30 and, following, a molecular sieve 32. The metal fiber filter 28 is in this case designed, in particular, as a fine filter with fiber filter mats having a decreasing fiber diameter of 40 μm to approximately 1 μm, so that, in particular, even penetrating fine aerosols with a particle size of less than 0.5 μm can be effectively retained. Additionally or alternatively, the Venturi scrubber 12 may also be followed by preferably double gravity-type drop separation with drop recirculation. The depressurization and activity retention system 4 of the nuclear plant 1 is designed for particularly reliable activity retention and, in particular, for a degree of separation of the washing device of even comparatively fine-grained aerosols with a particle size of less than 0.5 μm of 98% or above. For this purpose, the Venturi scrubber 12 and the throttle device 24 are deliberately coordinated with one another in terms of their dimensioning. This is based on the design aim that, in a response situation, the depressurization gas stream flows through the Venturi scrubber 12 at a particularly high flow velocity of more than 150 m/s, in particular of more than 200 m/s. To be precise, as became apparent, with such high flow velocities, a virtually abrupt rise in the separation rate can be achieved, and, in particular, even fine and the finest possible aerosol particles are bound in washing liquid droplets and are thus separated. A suitable selection of, in particular, the flow cross sections ensures in this case that, in virtually all the phases of an incident scenario, such a high flow velocity prevails in the Venturi scrubber 12. For this purpose, on the one hand, the throttle device 24 is designed, in a response situation, for operating essentially, that is to say at a system pressure above a limit pressure, in the range of critical depressurization. As a result, the sound velocity relevant for the flow medium is established in the gas stream flowing through the throttle device 24, independently of the system pressure prevailing in the safety containment 2. Owing to this flow velocity in the throttle device 24 which is independent of the system pressure in the safety containment 2, the volumetric throughput through the throttle device 24 is constant essentially independently of the system pressure prevailing in the safety containment 2, so that the volumetric throughput through the preceding Venturi scrubber 12 also correspondingly remains constant. To allow for sliding-pressure operation, that is to say direct action by the system pressure prevailing in the safety containment 2, the depressurization and activity retention system 4 is thus designed for ensuring that the flow passes at a correspondingly highly selected flow velocity through the Venturi scrubber 12 uniformly and virtually independently of the system pressure prevailing in the safety containment 2. This is also achieved in that the relevant pressure losses in the inflow line from the containment are minimized by the use of eccentric flaps with a low pressure loss and having zeta values <1, preferably <0.5. As can be seen in the enlarged illustration according to FIG. 2, the Venturi scrubber 12 comprises a plurality of Venturi tubes 16. The Venturi tubes 16 are in this case fed on the gas-stream side by a common supply system 40 connected on the inlet side to the pressure relief line 6. A comparatively large fraction of the Venturi tubes 16 is designed as what are known as long Venturi tubes which are arranged with their outlets 18 above the intended desired level 20 of the washing liquid W and which therefore issue directly into the gas space 22 in the manner of a “freely ejecting” arrangement. Furthermore, however, there is also provision for preventing soiling or an impairment of the operating behavior of the Venturi scrubber 12 due to settling or sedimentation, in that a comparatively small fraction, to be precise less than 10%, of the Venturi tubes 16 are oriented obliquely downward. An intensive circulation of the washing liquid W within the vessel 14 is achieved by means of these Venturi swirlers, so that sedimentation is reliably avoided. In particular, the Venturi tubes 16 designed as long Venturi tubes are designed for a comparatively high water load of the gas stream requiring treatment of more than 5, in particular more than 10, liters of washing liquid W per cubic meter of gas. For this purpose, an annular slit feed over the nozzle circumference at an opening angle of 30° to 45° is provided in the Venturi tubes 16 in the inlet region 42 for the washing liquid W. The dimensioning is in this case carried out in such a way that the ratio of the neck cross-sectional area determined at the constriction point 44 or neck, as it is known, of each Venturi tube 16 to the inlet area for the washing liquid W, determined at the annular slit feed, amounts to about 3:1. Moreover, the constriction point 44 is also that point at which the gas stream flowing through has its maximum flow velocity; consequently, the flow velocity taken into account for the design and coordination of the Venturi scrubber 12 and of the throttle device 24 is also determined at the constriction point 44. In the exemplary embodiment, the Venturi tubes 16 designed as long Venturi tubes are designed as round Venturi tubes with a neck width of less than 40 mm, so that, in the case of a passive intake and distribution of the washing liquid due to the underpressure generated by the medium flowing through, a feed of the washing liquid W into the core jet region inside the respective Venturi tube 16 is ensured. Furthermore, the Venturi tubes 16 have a ratio of height to neck width of more than 10. As may also be gathered, moreover, from the enlarged illustration according to FIG. 2, the throttle device 24 for drop separation is provided with an outflow tube 46 which issues on the outlet side into the washing liquid W. The throttle device 24 is itself connected on the outlet side to the pressure relief line 6. As may be seen, furthermore, from FIG. 1, to allow a particularly compact type of construction of the vessel 14, a multicomponent stock of washing liquid W is provided. On the one hand, washing liquid W in which the Venturi scrubber 12 is arranged is reserved in the vessel 14. Additionally, and to supplement this, however, the vessel 14 is connected on the washing-fluid side to a further washing liquid store 50 via a feed line 48. The washing liquid store 50 may be a receptacle which is designed specifically for this purpose and which is selected so as to lie at a geodetically suitable height for a reliable afterfeed of washing liquid W into the vessel 14, the desired level 20 of the washing liquid W in the vessel 14 being set by means of the height, set in the further washing liquid store 50, of the washing liquid W reserved there. Alternatively, however, the further washing liquid store 50 provided may also be a water tank provided in any case, such as, for example, a wastewater tank, a demineralized supply or the like, while the on-demand afterfeed of washing liquid W into the vessel 14 may take place via suitably selected gradients or, for example, by means of diaphragm pumps or compressed air. Furthermore, the vessel 14 is connected on the washing-fluid side to the interior of the safety containment 2 via a feedback line 52. This makes it possible to have, in the manner of a feedback, a recirculation of washing liquid W laden with airborne activities or with aerosols out of the vessel 14 into the safety containment 2. Consequently, by the constant or cyclic recirculation of washing liquid W laden in this way, the activity can be held in its entirety inside the safety containment 2 particularly reliably, so that the risk of a discharge into the surroundings is kept particularly low. Moreover, precisely because of such a recirculation of the washing liquid W, the decay heat imported via the retained activities can also be displaced consistently out of the vessel 14 back into the safety containment 2, so that the evaporation of washing liquid W in the vessel 14 is kept particularly low. Despite the recirculation of washing liquid W into the interior of the safety containment 2 and an afterfeed of washing liquid W out of the further washing liquid store 50, the overall consumption of washing liquid W which occurs can consequently be kept particularly low as a result of the avoidance of evaporation. As indicated by the dashed line 54, the feedback line 52 may be connected to the interior of the safety containment 2 via the pressure relief line 6. As illustrated in the enlargement of a detail in FIG. 3, the recirculation in this case takes place in the manner of a passive form in countercurrent to the gas stream emerging from the safety containment 2, no additional leadthrough through the safety containment 2 being required. In order in this case to ensure a sufficient feed pressure for the washing liquid W to be fed back, in the exemplary embodiment the vessel 14 together with the washing liquid W located in it is arranged at a sufficient geodetic height, to be precise about 10 m above the outlet point 56 of the pressure relief line 6 from the safety containment 2. Thus, solely due to the geodetic pressure in the water column in the feedback line 52, a sufficient feedback pressure for the washing liquid W into the safety containment 2 is ensured in the manner of a passive system. Alternatively, a cyclic feedback by the closing of the outlet fitting in the event of overpressure in the containment or the utilization of a separate small line of small subcritical cross section and corresponding action by pumps, for example a compressed-air diaphragm pump or a centrifugal pump, may also be provided. The components necessary for this purpose, for example a compressed-air reservoir 58, are illustrated diagrammatically in FIG. 1. For reliable iodine retention, the pH value in the washing liquid W in the vessel 14 is set at an alkaline value, in particular a value of more than 9. For this purpose, an on-demand addition of NaOH, other lyes and/or sodium thiosulfate takes place by intake via a jet pump located in the freshwater stream.
	description	This application claims the benefit of co-pending United States patent application entitled “TARGET SYSTEM FOR THE HANDLING OF A Cu-64 SOLID, LIQUID OR GASEOUS TARGET” filed Oct. 5, 2010 and assigned Ser. No. 12/898,087; co-pending United states patent application entitled “SOLID TARGET SYSTEM FOR THE HANDLING OF A Cu-64 TARGET” filed Jan. 30, 2006 and assigned Ser. No. 11/342,501; and United States provisional patent application filed Jan. 28, 2005 and assigned Ser. No. 60/648,147, which are incorporated by reference herein. 1. Field of the Invention The present invention relates to the field of positron emission tomography (PET). More particularly, this invention relates to a system and method for manually loading and remotely unloading a target disk into a proton beam. 2. Description of the Related Art Accelerators are commonly used to produce radionuclides for a variety of uses including Positron Emission Tomography (PET). PET is a noninvasive diagnostic imaging procedure that assesses the level of metabolic, biochemical, and functional activity and perfusion in various organ systems of the human body. PET provides information not available from traditional imaging technologies, such as Magnetic Resonance Imaging (MRI) and Computed Tomography (CT) which depict changes in anatomy rather than changes in physiology. Physiological activity provides a much earlier detection measure for certain forms of disease, cancer in particular, than do anatomical changes over time. Typically, an accelerator produces radionuclides by accelerating a particle beam and bombarding a target material with the accelerated beam thereby producing radionuclides. The type of radionuclides produced are determined by the target material and particle beam used. Low or medium energy charged-particle accelerators typically produce radionuclides having a short half life. Radionuclides such as copper-64 or 64Cu have a longer half life than the conventional radionuclides typically used. Specifically, copper-64 is the cyclotron-produced PET isotope of copper. This isotope undergoes a special type of radioactive decay, whereby its nuclei emit positrons that travel only a few millimeters in tissue before colliding with electrons, converting their total mass into two photons of energy. The photons are displaced at 180 degrees from each other and can be detected simultaneously as “coincident” photons on opposite sides of the body. However, copper-64 is not easily producible as is shown in U.S. Pat. No. 6,011,825 which is incorporated herein in its entirety by reference. The production of copper-64 requires the irradiation of a solid target rather than a liquid or gaseous target that conventional accelerators are capable of handling. The combination of gold with plated enriched nickel can be used to produce copper-64. Other combinations of metals can also be used to provide copper-64. In addition, the combination of metals can take the form of pellets, foil or coin. There is a need for a target holder for loading and unloading a solid target to produce a radionuclide. There is also a need for a target holder that can accommodate a solid as well as a liquid and gas target cost effectively. There is a further need for a target holder that has a service position and an irradiation position An object of the present invention is to provide a solid target handling system for manually loading and remotely unloading a target disk into a proton beam. Another object of the present invention is to provide a target handling system that can efficiently and cost effectively accommodate a solid target, a liquid target and a gas target. An aspect of the present invention provides a system and method for a system for accommodating a solid target in an accelerator. The system and method includes a target changer having at least one port for accommodating the solid target, an insert for receiving the solid target in the target changer, a piston for providing a vacuum and a cooling system for the solid target, a cylinder for displacing the piston in one of three positions; and a bracket for securing the insert, piston and cylinder to the target changer. Another aspect of the present invention also provides a system and method for accommodating a solid target, a liquid target and a gaseous target mounted on an accelerator. The system and method provide a target changer having four ports, two of which are service positions, an insert for receiving the solid target in the target changer, a piston for providing a vacuum and a cooling system for the solid target, a cylinder for displacing the piston in one of three positions; and a bracket for securing the insert, piston and cylinder to the target changer in one of the ports. A further aspect of the present invention provides for the target changer being rotated from a first position to a second position, wherein the first position comprises a service/removal position and the second position comprises a beam position for irraditiating the solid target. Throughout the figures, like symbols and numbers are used throughout. The solid target handling system 10 is configured with several criteria. First, the system 10 is received and operates in a conventional shield envelope (not shown). The system 10 is mounted to a conventional exiting target changer hub 24 as shown and described in U.S. Pat. No. 5,608,224 which is incorporated herein by reference in its entirety, and interfaces to an existing cooling arrangement. The hub 24 also mounts to an adjustable back plate for alignment to a beam. The beam has a range of about 5 MeV to about 25 MeV. Preferably, the beam has energies at about 11 MeV. FIGS. 1-4 show the above described components and assembly. More specifically, FIG. 1 illustrates a target changing system in accordance with an embodiment of the present invention. FIG. 2 illustrates an elevation view, in section, of the target assembly in accordance with an embodiment of the present invention. FIG. 3 illustrates the target changer 2 having four ports in accordance with an embodiment of the present invention. FIG. 4 illustrates various components of the target system in accordance with an embodiment of the present invention. The basic operation of the target changer interfaces with a conventional accelerator control system (not shown). The unloading of the system 10 is controlled by a remote controller not shown), positioned outside the shield, with operational logic. The system 10 accommodates all conventional eclipse style targets in two ports, and accommodates a solid target in another two ports. The system 10 comprises a target changer 2, an insert 4, a piston 6, a shaft 22, a cylinder 8, a bracket 12 and a feed slot 14 as shown in FIG. 1. The insert 4 has an o-ring 16, a first opening 7, a second opening 9 and a cavity (not shown) providing a pass through between the first opening 7 and the second opening 9. The first and second openings of the insert 4 can be the same size; the first opening can be larger than the second opening or vice versa. The insert 4 also includes a slot 3. The slot 3 is positioned and arranged to allow a target to fall through from the feed slot 14. The piston 6 has a tab 5 and an o-ring 20. The feed slot 14 is located within the target changer 2. FIGS. 1, 2 and 3 together further show target changer 2 having a first port 26 for accommodating the insert 4, the piston 6, the shaft 22, the cylinder 8, and the bracket 12 all of which comprise subsystem 11. Target changer 2 also includes a third port 28 disposed about 180 degrees from the first port 26. It should be appreciated by those skilled in the art that the positions of the first port 26 and third position port 28 can vary from 180 degrees without departing from the scope of the present invention. For example, the first port 26 and the third port 28 can be 90 degrees apart without varying from the scope of the present invention. As shown in the combination of FIGS. 1-5, first port 26 is in the service/removal position. Third port 28 is in the beam position. The target changer 2 is rotated so that the first port 26 is displaced from a service position to a beam position. Second port 30 and fourth port 32 can accommodate conventional liquid and gas targets. In an embodiment of the present invention, target changer 2 comprises only first port 26 and third port 28. In another embodiment of the present invention, target changer 2 comprises a plurality of first ports 26 and a plurality of third ports 28. This will enable a plurality of solid targets to be accommodated and produce substantial amounts of radionuclides in a short amount of time. In operation, the solid target is manually loaded in the first port 26 or the service position of the target changer 2. The target extraction mechanism is then attached to the target via computer control. The target is then rotated into the beam position and bombarded for the desired time and current. The target is then rotated back to the service port and unloaded. The unloading process includes the following steps. First, the solid target is rotated to the service/removal position. The first port 26 vacuum line 40 is then vented. The cooling water valve 36 is closed, and then opened to drain. An air flush valve 42 is opened to remove all trapped water from the cooling lines. The target removal mechanism is initialized and the target is extracted from the insert 4. The target falls out of the device and to the floor of the accelerator pit aided by gravity. The fall is within a track (not shown) to control speed and location. The target changer 2 is then available to manually load another solid target. FIG. 4 illustrates an exemplary target 34. Target 34 is a solid target and preferably comprises a combination of enriched nickel and gold sufficient to provide copper-64. The piston 6 fits within the insert 4 and channels cooling water to the solid target via perforations 44 (See FIGS. 1, 4 and 6). The insert serves as the vacuum seal between the target and the accelerator. The piston has three positions within the insert. A load, extended and extraction position. The load position is such that the tab 5 on the piston extends into the slot 3 of the insert 4 preventing the target from continuing to fall out of the feed slot 14 where it exits the target changer. Specifically, the tab 5 (see mark up to FIG. 4) stops the target disk as it falls into the target changer 2 and positions the target in the center of the beam. In the extraction position the piston 6 is extracted in the insert 4. It should be appreciated by those skilled in the art that the extraction position can comprise a location where the piston 6 is still in the insert 4 but the tab 5 is not blocking feed slot 14. The three positions of the piston are controlled by a pneumatic cylinder 8 manufactured by Bimba. The cylinder is held in position by the bracket 12, which is connected to first port 26 via screws and precisely positions the cylinder 8 so that the stroke lengths are as needed. The displacement of shaft 22 which is connected to cylinder 8 at one end and piston 6 at a distant end causes piston 6 to move in a lateral direction. In an embodiment of the present invention, the system 10 is configured to accommodate a solid target having a range between 0.5 mm to 5 mm thickness and 10 mm to 35 mm in diameter. Preferably the target disk has 2 mm thickness and 25 mm diameter. The solid target preferably has a thermal conductivity greater than 2200 BTU-in/hr-Ft2-° F. In accordance with an embodiment of the present invention, system 10 operates in the following manner. When first port 26 is in the service position, the target 34 is dropped either manually or remotely into the feed slot 14 of the target changer 2. The feed slot 14 was formed via a rectangular slot that was burned into the target changer 2 via EDM. The feed slot 14 allows the target disk to fall by gravity into the insert 4. The target enters the insert 4 via the insert slot 3 and is prevented from passing through the insert 4 by the piston tab 5 because the piston 6 is in the load position. Air is removed via air inlet 40 compressing the target against the o-ring 16 of insert 4. The piston is placed in an extended position compressing the target against O-ring 20 of the piston 6. The port 26 is rotated by the hub 24 from a service position to a beam position where the target is irradiated for a predetermined period by a beam having a predetermined energy. An exemplary predetermined time period can be 2 hours of 40 uA operation for the accelerator. In an embodiment of the present invention, the rotation can be clockwise. In another embodiment of the present invention, the rotation can be counter clockwise. Water is input via inlet 36 and the perforations 44 of the piston 6 to maintain the temperature of the target below a predetermined threshold temperature so that the target does not melt. Water is removed via outlet 38. The target changer 2 is rotated clockwise so that first port 26 is positioned to be in a removal position. In another embodiment of the present invention rather than continuing forward in a clockwise direction, the target changer 2 is rotated in a counter clockwise position. In the removal position, air is provided to port 26 via inlet 42, the piston 6 is placed in an extracted position causing the target to fall through slot 3 of the insert 4 via gravity out of the target changer 2 where the target is automatically unloaded and interfaces with a customer supplied transport system. The insert is designed to fit within the target changer. It functions to position the 25 mm diameter solid target in the larger target slot. It provides cooling water and vacuum seals. It also has integral tabs to strip the target disk from the piston during extraction. The beam position compresses the target disk between two face seal O-rings for vacuum seal. The extract position pulls the piston back within the insert and allows the target disk to fall into the exit feed slot. The operation of the target assembly of the present invention is illustrated in FIG. 5. Target Cooling: The target disk is cooled by water jets normal to its non-beam side surface. The water is routed through the insert as indicated in FIG. 6. The target disk is cooled by conduction through the disk and convection from the disk into the cooling water. Conduction is calculated by Fowlers Law: q = KA ⁢ ⅆ t ⅆ l . Since the heat transmission is steady and the K and L are constant, this becomes: q = KA ⁢ ⁢ Δ ⁢ ⁢ T L ,where: q=heat input; K thermal conductivity of material; A=area of heat conduction; ΔT=(T2−T1); and L=thickness of target disk. In the instant case, where: q=10.5 MeV×60 uA=630 W=2150 btu/hr; K=2200 btu-in/hr-ft2-° F. (for gold); A=0.00136 ft2; and L=2 min=0.079 in,then: ΔT=57° F. To estimate the value for “h”, the coefficient of heat transfer used in the following equations are used: H=Nu(Kwater)/L; Nu=0.228 Re0.731Pr0.33 Re=VLρ/μ; and Pr=Cpμ/Kwater, where: Nμ=Nusselt number; Re=Reynolds number; Pr=Prandlt number; Kwater=thermal conductivity of water=0.58 W/m K; L=length of flow=0.019 m; P=density of water=1000 kg/m3; M=viscosity of water=0.00114 kg/m-s; Cp=specific heat of water=4180 KJ/kg-K; and V=velocity of flow=4.3 m/s. From this, the results yield: Pr=8.2; Re=7.2×104; Nu=1627; and H=49,667 W/k=8741 btu/hr-ft2-° F. Convection is calculated by Newton's Law of Cooling for forced convection. q=hAΔTwhere: q=heat input; h=coefficient of heat transfer; A=area of heat convection; and ΔT=(Twall−Twater). In the instant case, where: q=10.5 MeV×60 uA=630 W=2150 btu/hr; h=8741 btu/hr-ft2-° F.; and A=0.00136 ft2,then: ΔT=180° F. The results show that where the temperature of the cooling water is 45° F., the temperature of the wall on the cooling water side is 225° F. and the temperature of the wall on the beam side is 282° F. While the present invention has been illustrated by description of several embodiments and while the illustrative embodiments have been described in considerable detail, it is not the intention of the applicant to restrict or in any way limit the scope of the appended claims to such detail. Additional advantages and modifications will readily appear to those skilled in the art. The invention in its broader aspects is therefore not limited to the specific details, representative apparatus and methods, and illustrative examples shown and described. Accordingly, departures may be made from such details without departing from the spirit or scope of applicants general inventive concept.
	abstract	Exemplary embodiments provide automated nuclear fission reactors and methods for their operation. Exemplary embodiments and aspects include, without limitation, re-use of nuclear fission fuel, alternate fuels and fuel geometries, modular fuel cores, fast fluid cooling, variable burn-up, programmable nuclear thermostats, fast flux irradiation, temperature-driven surface area/volume ratio neutron absorption, low coolant temperature cores, refueling, and the like.
056688431	summary	CROSS-REFERENCE TO RELATED APPLICATION This application is a Continuation of International Application Ser. No. PCT/DE95/00066, filed Jan. 18, 1995. BACKGROUND OF THE INVENTION FIELD OF THE INVENTION The invention relates to a storage cage for the storage and transport of a plurality of fuel assemblies, in particular fuel assemblies which have been used in a nuclear reactor for the release of energy by nuclear fission in fissionable material that is contained in the fuel assemblies. Both the storage and the possibly necessary transport of fuel assemblies, in particular after the fuel assemblies have been used in a nuclear reactor, necessitate an abundance of measures arising from reasons of radiation protection, a need to dissipate heat originating from radioactive decays in the fuel assemblies and a requirement of preventing the formation of an accumulation of fissionable material in which an automatic chain reaction of nuclear fissions could occur. The first-mentioned reason necessitates a careful shielding of an irradiated fuel assembly, the second-mentioned reason demands special measures for heat dissipation and the third-mentioned reason necessitates that a multiplicity of fuel assemblies always be gathered together in a narrow space only with an appropriate quantity of neutron-absorbing material. German Patent DE 32 21 810 C2 discloses a device for the storage of irradiated fuel assemblies. The device is to be disposed in a nuclear power station in the vicinity of the nuclear reactor and inside a guard shield surrounding the nuclear reactor. The device includes a storage pond filled with light water and a rack which is disposed in the storage pond and in which the irradiated fuel assemblies can be stored. The rack has a baseplate, on which casings made of boron-containing steel are mounted in a configuration in the manner of a honeycomb. Each casing serves for receiving a fuel assembly. The shielding of the fuel assemblies is guaranteed, in particular, through the use of water. The water also serves, together with the material of the casings, for the absorption of neutrons, in order to reliably rule out the buildup of a chain reaction. A method and a storage device for the treatment of spent fuel assemblies from nuclear power stations are disclosed in German Published, Non-Prosecuted Patent Application DE 34 00 929 A1. According to that publication, fuel assemblies which are to be removed from a nuclear power station are sealingly enclosed in cans inside the guard shield of the nuclear reactor. Only sealingly closed and, where appropriate, externally decontaminated cans are brought out of the guard shield and stored in an external storage device which is equipped with air cooling for the dissipation of heat energy from the fuel assemblies. German Published, Non-Prosecuted Patent Application DE 28 40 594 A, like German Patent DE 32 21 810 C, relates to a storage rack for fuel assemblies to be placed in a water pond. The storage rack is composed of rack parts which are disposed next to one another and, where appropriate, also one above the other. A rack part has projections and recesses at the edge of its cross section, which can be fitted together with corresponding recesses or projections in adjacent rack parts and which, in particular, form supporting surfaces between the rack parts. The rack formed from the rack parts, which can be stacked, in particular, in the manner of beer crates, can absorb not only vertically acting, but also horizontally acting forces. Published European Patent Application 0 385 186 A1 describes a fuel assembly storage rack with an outer frame which has transport lugs at its upper end and which includes a horizontal support plate at its lower end. Vertical tubes having a baseplate are disposed in the outer frame and through the use of which they are fastened to the support plate. Vertical wall surfaces of the tubes run parallel to one another and to the outer frame. In order to ensure the passage of cooling water, the baseplates or the vertical wall surfaces have perforations which are matched with corresponding perforations in the outer frame. The transport lugs serve for lifting the storage rack through the use of a lifting appliance into a fuel assembly storage pond which is provided and for positioning it there. U.S. Pat. No. 4,960,560 specifies a fuel assembly storage rack with a baseplate and cells being fastened to the base-plate, being extended perpendicularly to the latter and being intended for receiving fuel assemblies. The baseplate contains an aperture for each cell, for the entry of cooling liquid. In order to lift the storage rack, after its assembly, into the intended location in a fuel assembly storage pond, some of the apertures are formed in such a way that an appliance for lifting the storage rack can engage into them. Through the use of the appliance, the storage rack and the fuel assemblies which are disposed therein solely vertically relative to the baseplate, can be lifted into different positions within the fuel assembly storage pond. The known possibilities for the storage and transport of fuel assemblies always involve the need, where appropriate, to reload fuel assemblies individually from a first storage device into a second storage device, for example from a rack into a transport container. In view of the special requirements to be placed on the handling of such fuel assemblies, that means that handling should always take place only in a specially shielded environment. That also results in a very high outlay which is extremely undesirable, not the least for reasons of radiation protection, since individual fuel assemblies have to be moved for each reloading operation. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a storage cage for storage and transport of fuel assemblies, which overcomes the hereinafore-mentioned disadvantages of the heretofore-known devices of this general type and which provides a possibility of substantially reducing outlay during transfer of fuel assemblies in comparison with the heretofore-known devices. With the foregoing and other objects in view there is provided, in accordance with the invention, a storage cage for the storage and transport of a plurality of fuel assemblies, comprising a baseplate having apertures formed therein; a plurality of casings each having an interior and each standing on the baseplate for receiving and storing a fuel assembly; each of the casings being associated with at least one of the apertures leading into the interior of the casing for supplying and discharging cooling liquid; each of the casings having a plurality of slots formed therein for supplying and discharging cooling gas; and at least one supporting wall forming a load-bearing part with the baseplate, the casings being fastened to the at least one supporting wall. On one hand, the storage cage can be disposed with a plurality of identical storage cages provided with irradiated fuel assemblies, in a water-filled storage pond, in order to store fuel assemblies directly next to a nuclear reactor. In order to ensure that the storage cage can be taken out of the storage pond easily, the baseplate has apertures formed therein, through which a cooling medium, such as water, can flow off out of the casings. Furthermore, the apertures serve for generating a natural circulation of the water along the fuel assemblies for the cooling thereof. Moreover, the storage cage together with the fuel assemblies can be introduced into a transport container and be anchored therein, in order to thereby allow easy transport of the fuel assemblies as a whole. For this purpose in particular, the storage cage has the supporting wall which absorbs all of the loads to be expected, even in the event that the storage cage does not stand on the baseplate for transport, but lies with the baseplate oriented vertically. Each casing is provided with a plurality of slots to ensure that the gas cooling of the fuel assemblies which is conventionally carried out in a transport container becomes possible. This affords a possibility for the storage, transfer and transport of fuel assemblies which largely avoids the handling of individual fuel assemblies. After the fuel assemblies have been introduced into the storage cage, it is only ever necessary for the storage cage as a whole to be moved. In accordance with another feature of the invention, the storage cage is constructed, particularly by the provision of a suitable anchor device, in such a way that it can be connected to a plurality of identical storage cages, with all of the baseplates being oriented horizontally, to form a storage rack. An anchor device of this type can, for example, be noses which have bores formed therein and which can be screwed through the use of bolts or the like to adjacent storage cages having corresponding noses. Configurations of catches and hooks, through the use of which mutually adjacent storage cages can be hooked in one another, are also possible. In accordance with a further feature of the invention, the storage cage can be stacked with another identical storage cage, with the baseplates being oriented horizontally. In accordance with an added feature of the invention, the storage cage is constructed in such a way that it can be anchored in a sealingly closeable associated container, particularly for transport purposes. In accordance with an additional feature of the invention, the casings of the storage cage are formed of a neutron-absorbing material, preferably of a boron-containing material, in particular boron-containing steel. Since boron-containing steel is relatively brittle, it usually requires an additional supporting device in order to ensure a sufficient load-bearing capacity of the storage cage. The supporting wall serves this purpose. In accordance with yet another feature of the invention, the baseplate and the supporting wall are formed of a rust-resistant steel. In accordance with yet a further feature of the invention, each casing of the storage cage is shrouded by an associated supporting wall. Such a storage cage does not have any exposed surfaces being formed of brittle material, which is particularly advantageous with respect to the load-bearing capacity of the storage cage. In accordance with a concomitant feature of the invention, the casings of the storage cage form an approximately cylindrical configuration, so that conventionally constructed, that is to say essentially cylindrical containers, can be used for transporting the storage cage. It goes without saying that the storage cage is provided, according to the requirements of the particular individual instance, with fastening elements for securing suitable lifting appliances. The fastening elements are furthermore advantageously constructed in such a way that they serve as rests and fastening points for a further storage cage which is to be placed onto the first-mentioned storage cage. It is also advantageous if the fastening elements of this type are constructed in such a way that they can serve for fastening the cage in a corresponding transport container. For this purpose, the storage cage is preferably also provided with lateral supports, in order to support the storage cage with negligible play in a container for transport, and where appropriate a horizontal storage of the container and of the fuel assemblies is necessary during transport. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a storage cage for storage and transport of fuel assemblies, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims.

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