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047972327	description	The examples which follow will illustrate the invention. Example 1: 1st case The solutions On the laboratory scale, a solution of FP was simulated using a typical composition of a real solution of FP in the following manner: ______________________________________ Corresponding Quantity quantity of oxide Product used (g) (g) ______________________________________ 1- Al(NO.sub.3).sub.3.9H.sub.2 O 117.6 15.9 Fe(NO.sub.3).sub.3.9H.sub.2 O 146.7 29 Ni(NO.sub.3).sub.2.6H.sub.2 O 19.4 5 Cr(NO.sub.3).sub.2.9H.sub.2 O 26.3 5 Na.sub.4 P.sub.2 O.sub.7.10H.sub.2 O 9.4 5.6 NaNO.sub.3 103.6 37.7 2- Sr(NO.sub.3).sub.2 6.7 3.2 CsNO.sub.3 15.2 10.9 Ba(NO.sub.3).sub.2 9.7 5.6 ZrO(NO.sub.3).sub.2.2H.sub.2 O 34.7 15.9 Na.sub.2 MoO.sub.4.2H.sub.2 O 26.4 22.5 Co(NO.sub.3).sub.2.6H.sub.2 O 5.8 1.4 Mn(NO.sub.3).sub.2.4H.sub.2 O 27.7 9.5 Ni(NO.sub.3).sub.2.6H.sub.2 O 18.3 4.6 Y(NO.sub.3).sub.3.4H.sub.2 O 5.5 1.7 La(NO.sub.3).sub.3.6H.sub.2 O 23.7 8.8 Ce(NO.sub.3).sub.3.6H.sub.2 O 24.9 9.3 Pr(NO.sub.3).sub.3.4H.sub.2 O 10.6 4.3 Nd(NO.sub.3).sub.3.6H.sub.2 O 39.6 15.1 ZrO.sub. 2 4.6 4.6 Mo 3.5 5.3 U.sub.3 O.sub.8 8.8 8.5 ______________________________________ Group 1 represents the inactive components of the solution of FP and group 2 simulates the active components of this same solution and the insoluble materials. ZrO.sub.2 and Mo remain solid; they simulate the insoluble materials suspended in the solution. The total quantity of water added is 2972 g. The simulated solution of FP has a pH of 1.3. The composition of the final glass to be obtained is: ______________________________________ Composition of the glass introduced via ______________________________________ SiO.sub.2 45.5% Ludox B.sub.2 O.sub.3 14% Solution of ATB Al.sub.2 O.sub.3 4.9% Solution of adjuvant and simulated solution of FP Na.sub.2 O 9.8% Solution of adjuvant and simulated solution of FP ZnO 2.5% Solution of adjuvant and simulated solution of FP CaO 4.1% Solution of adjuvant and simulated solution of FP Li.sub.2 O 2% Solution of adjuvant and simulated solution of FP Active oxides 13.2% Simulated solution of FP Fe.sub.2 O.sub.3 2.9% " NiO 0.4% " Cr.sub.2 O.sub.3 0.5% " P.sub.2 O.sub.5 0.3% " ______________________________________ In the percentage composition shown, it is necessary to allow for the presence of Na and Ni in the active oxides (group 2 of the solution of FP defined above). Thus, the solution of the vitrification adjuvant is prepared according to the composition of the glass to be obtained and the composition of the solution of waste to be treated. For this example, the solution of vitrification adjuvant is prepared as follows: ______________________________________ Corresponding Quantity quantity of oxide Product used (g) (g) ______________________________________ Al(NO.sub.3).sub.3.9H.sub.2 O 243.6 33.1 NaNO.sub.3 148.4 54.1 Zn(NO.sub.3).sub.2.6H.sub.2 O 91.4 25 Ca(NO.sub.3).sub.2.4H.sub.2 O 170.1 40.4 LiNO.sub.3 91.4 19.8 ______________________________________ Each of the compounds is dissolved in the minimum quantity of water, i.e. a total of 640 g of water at 65.degree. C.; pH: 0.6. The precursor is Ludox AS40: 40% SiO.sub.2 /60% H.sub.2 O; .phi. of the particles: 21 nm; d.sub.25.degree. C. : 1.30; pH: 9.3; used at ambient temperature. The ATB solution is 265.2 g of (NH.sub.4).sub.2 0.2B.sub.2 O.sub.3.4H.sub.2 O dissolved in 663 g of water at 65.degree. C.; pH: 9.2. The device The device used is a conventional turbine having a mixing zone of small volume, in which a propeller with several blades rotates so as to effect mixing at a high rate of shear. It rotates at 2000 rpm in this example. The turbine used for the tests is manufactured by the Company STERMA, the mixing zone has a volume of 1 cm.sup.3 and the thickness of the stirred layer is of the order of mm. The procedure The solutions arrive at the turbine separately and simultaneously: ______________________________________ Flow rate Composition of pH T.degree. at T.degree. the solution ______________________________________ Ludox 9.3 20.degree. C. 12 kg/h 40% of SiO.sub.2 Ammonium tetra- 9.2 65.degree. C. 9.9 kg/h 21% of anhydrous borate.4H.sub.2 O (ATB) salt, i.e. 15% of B.sub.2 O.sub.3 Solution of 0.6 65.degree. C. 14.7 kg/h 40% of anhydrous vitrification salt, i.e. 12% adjuvant of oxides ______________________________________ 36.5 kg/h of borosilicate matrix are thus prepared. 1.7 kg are spread over a plate with an average thickness of 2 cm and then placed in an oven at 100.degree.-105.degree. C. for 48 hours; 0.6 kg of dry matrix is obtained. 1.6 l of simulated solution of FP are placed in a 3 l container equipped with a rotating mechanical stirrer; the dried matrix is poured in uniformly, with stirring. The mixture obtained is stirred for about 30 min and then dried at 100.degree.-105.degree. C. in an oven on a plate, calcined for 2 h at 400.degree. C. and finally melted for 5 h at 1050.degree. C. The glass obtained (0.5 kg) satisfies the criteria of acceptability. In the tests, a glass of good quality was defined as being a homogeneous glass having no unmelted regions and no bubbles and also showing no traces of molybdate on the surface. The molybdate originating from the solutions of FP actually presents a major problem: part of the active Mo tends to separate out from the solution and deposit, so this phase is not completely dispersed in the mixture and hence is not totally included in the gelled solution. Furthermore, when it diffuses poorly, the molybdenum appears on the surface of the glass in the form of visible yellow traces of molybdate, which are considered to be an indication of inferior quality glass. Chemical analysis of the glass obtained further shows that the components are not volatilized in practice, so it can be considered that the composition of the mixtures (borosilicate matrix, then matrix+waste) virtually corresponds to that of the final glass. Example 2: 2nd case Test 1 3.7 kg of the borosilicate matrix coming from the turbine (preparation according to Example 1) are dried for 20 hours at 100.degree.-105.degree. C. on plates in an oven. The dried matrix is then placed in a furnace in which the temperature is gradually raised to 350.degree. C. over 2 hours, and calcination is carried out for 2 h at 350.degree. C. The product obtained is friable and is in the form of fragments of a few mm in diameter (on average 2-3 mm). The calcined matrix (1 kg) is ground (.perspectiveto.300-400 .mu.) and dispersed in the solution of FP (3 kg), simply with stirring (magnetic stirrer, 30-45 min). The mixture is calcined for 4 h at 400.degree. C. after being heated for 34 h at 120.degree. C., and is then melted at 1125.degree. C. 40 min in the introduction zone and 1 h in the refining zone lead to a glass of good quality. Test 2: This test relates to the treatment of the soda effluent used for washing, which is subsequently acidified. At present, in the vitrification (AVM) process based on the oxides, it is not easy to treat this effluent on its own. This AVM process actually uses the vitrification adjuvant in the form of a solid glass frit, a known composition being: ______________________________________ SiO.sub.2 55-60% by weight B.sub.2 O.sub.3 16-18 by weight Al.sub.2 O.sub.3 6-7 by weight Na.sub.2 O 6-7 by weight CaO 4.5-6 by weight ZnO 2.5-3.5 by weight Li.sub.2 O 2-3 by weight ______________________________________ If this composition were used to vitrify the soda effluent, the glass obtained would be very rich in sodium. One might consider reducing the level of sodium in the glass frit, even to zero, so that the final glass (frit+calcinate of soda effluent) has an acceptable sodium level (9 to 11% by weight). However, one is then faced with the difficulty of producing and melting a glass which is poor in sodium (and consequently richer in silica). The present invention makes it possible to produce, with the soda effluent, a borosilicate glass having a composition similar to that which proves totally satisfactory in the AVM process. Moreover, the refining temperature can be considerably lowered or the refining times shortened. For tests, a soda solution was therefore simulated using 100 g of Na.sub.2 CO.sub.3 in one liter of water. The ATB solution contains 312 g/l of ATB.4H.sub.2 O. The boric acid solution contains 130 g/l (6.5% of B.sub.2 O.sub.3)-pH=2.7. To obtain a glass having a composition similar to that obtained by the AVM process, the following solution of vitrification adjuvant is prepared (amounts are per liter of aqueous solution): ______________________________________ Al(NO.sub.3).sub.3.9H.sub.2 O 209.0 g Ca(NO.sub.3).sub.2.3H.sub.2 O 98.5 g LiNO.sub.3 53.7 g Zn(NO.sub.3).sub.2.6H.sub.2 O 49.7 g Fe(NO.sub.3).sub.3.6H.sub.2 O 73.5 g Mn(NO.sub.3).sub.3.6H.sub.2 O 18.2 g Ba(NO.sub.3).sub.2 5.5 g Co(NO.sub.3).sub.2.6H.sub.2 O 11.3 g Sr(NO.sub.3).sub.2 4.1 g CsNO.sub.3 8.0 g Y(NO.sub.3).sub.3.4H.sub.2 O 71.0 g Na.sub.2 MoO.sub.4.2H.sub.2 O 16.6 g Monoammonium phosphate 2.8 g ______________________________________ The components Fe, Mn . . . phosphate were introduced into this solution so as to give a final glass with a composition similar to that given in the previous examples. On the other hand, Aerosil.RTM., marketed by the firm DEGUSSA, will be used instead of Ludox AS40 as the gel precursor. The gel precursor is formed by pouring the Aerosil gradually, with stirring, into water acidified with 3 N HNO.sub.3 (pH: 2.5), so as to give a solution containing 150 g of silica per liter. 3 diaphragm pumps are provided, which have been adjusted beforehand to give the desired flow rates. The following solutions are pumped simultaneously into a high-speed mixer (capacity: 1.5 liters) at the indicated flow rates and temperatures. The set flow rates are: ATB solution . . . 0.57 l/h at 65.degree. C., or alternatively PA0 H.sub.3 BO.sub.3 solution . . . 1.25 l/h at 65.degree. C. PA0 Adjuvant solution . . . 1.15 l/h at 65.degree. C. PA0 Aerosil solution . . . 2 l/h at 20.degree. C. PA0 ATB solution containing 15% of B.sub.2 O.sub.3 at 0.75 l/h, or alternating H.sub.3 BO.sub.3 solution containing 6.5% of B.sub.2 O.sub.3 at 1.7 l/h PA0 Aerosil solution containing 150 g of SiO.sub.2 /l at 1.3 l/h PA0 Adjuvant solution containing 12% of oxides at 0.75 l/h The borosilicate matrix, obtained in the form of a gelled solution, is dried for 24 h at 105.degree. C. and then calcined for 3 h at 350.degree. C.. The solid particles taken from the furnace have a large specific surface area which varies from test to test but is always close to 50 M.sup.2 /g. After cooling, these particles are poured into the effluent to be treated and the mixture is stirred for 2 h. A gelatinous mass is formed, which is dried at 105.degree., calcined at 400.degree. C. and finally melted at 1150.degree. C. Chemical analysis gives the following average composition: ______________________________________ SiO.sub.2 45.6% B.sub.2 O.sub.3 14% Na.sub.2 O 10% Al.sub.2 O.sub.3 4.9% CaO 4% Li.sub.2 O 2% Fe.sub.2 O.sub.3 2.9% MnO.sub.2 0.95% BaO 0.55% CaO 0.5% Cs.sub.2 O 1% SrO 0.35% Y.sub.2 O.sub.3 4% MoO.sub.3 2% P.sub.2 O.sub.5 0.3% ______________________________________ Example 3: 3rd case Test 1 The following are introduced simultaneously into a 2 l mixer in 1/2 h: 1.4 kg of mixture are obtained; this is dried at 100.degree.-105.degree. in an oven on a plate, then calcined for 3 h at 350.degree. and finally melted. 320 g of this inactive calcined matrix are added to 135 g of a calcinate of FP and the two are roughly mixed. A melting time of 2 h at 1100.degree. C. is required to give 300 g of a glass of the desired composition (that of Examples 1 and 2). This example shows that it is possible to prepare a calcined gel having the same composition as the glass frit used in the AVM process. Test 2 Here it is desired to vitrify a mixture of solution of FP+soda effluent. This is done by preparing a calcined matrix having a composition similar to the glass frit of the AVM process, except for the sodium: the sodium oxide level is reduced from 7% to 2.6%. The solution of vitrification adjuvant will have the following composition: ______________________________________ Corresponding Product used Quantity in grams weight of oxide ______________________________________ NaNO.sub.3 55.1 20.1 Al(NO.sub.3).sub.3.9H.sub.2 O 243.6 33.1 Zn(NO.sub.3).sub.2.6H.sub.2 O 91.4 25.0 Ca(NO.sub.3).sub.2.4H.sub.2 O 170.1 40.4 LiNO.sub.3 91.4 19.8 ZrO.(NO.sub.3).sub.2.2H.sub.2 O 11.7 5.4 ______________________________________ The matrix will be completed using: as the source of silicia: Ludox AS40 PA1 as the source of boron: a boric acid solution containing 130.5 g per 1000 g of water, kept at 60.degree. C. PA1 solution of vitrification adjuvant: 5 kg/h PA1 solution of Ludox: 9.5 kg/h PA1 solution of boric acid: 5.8 kg/h PA1 (1) Ludox AS40 at 65.degree. C., 1150 g PA1 (2) ATB.9H.sub.2 O at 65.degree. C. in solution at the saturation limit (about 40 g/100 g of water), 312 g PA1 (3) a solution of vitrification adjuvant practically saturated with lithium and sodium nitrates at 65.degree. C., containing 225 g of NaNO.sub.3 and 87.5 g of LiNO.sub.3 in 250 g of water. The following flow rates are delivered simultaneously to the turbine with three pumps: Practically 20 kg of a gel are recovered in one hour; this is dried on a plate in an oven at 100.degree.-105.degree. C. and then calcined at 400.degree. C. (with gradual increase in temperature and a plateau at 200.degree. C.). This gives a solid mass composed of irregular pieces of a few cm.sup.3. These are ground to a uniform size and sieved with a 2.5 mm mesh. Analysis of this calcined product gives: ______________________________________ SiO.sub.2 61.6 (% by weight) B.sub.2 O.sub.3 19 (% by weight) Na.sub.2 O 2.7 (% by weight) Al.sub.2 O.sub.3 4.5 (% by weight) ZnO 3.4 (% by weight) CaO 5.5 (% by weight) Li.sub.2 O 0.75 (% by weight) ______________________________________ This analysis can be seen to be very similar to the formulations of the typical frit used in the AVM process as regards all the constituents except sodium. The ratio of silica to boric oxide is equal to 3.244 in the theoretical formula and 3.242 in the calcined gel. The ratio of silicia to alumina is equal to 13.75 in the theoretical formulation and 13.69 in the calcined gel. By contrast, the ratio of silica to sodium is equal to 8.407 in the theoretical formulation and 22.82 in the calcined gel. The sodium level is 7% in the theoretical formula and 2.7% in the calcined gel. Thus, a mixture of solution of FP+soda effluent can be treated by vitrification while preserving a normal sodium level for the final glass, as shown in the remainder of the example. 2500 g of a solution of sodium nitrate containing 100 g/kg, simulating the soda effluent, are added to 10 liters of the solution simulating the FP (as described in Example 1). (Sodium nitrate is used because the solution simulating the FP contains no free nitric acid, which is unrealistic.) The mixture is dried at 105.degree. C. on a plate in an oven and then calcined at 400.degree. C. in a small furnace to give a powder consisting of grains of a few millimeters, which represent the calcinate of (FP+soda effluent) and which we will refer to as the calcinate. 375 g of the said calcinate are carefully mixed dry with 1000 g of the calcined gel. The mixture is introduced in several portions into a crucible placed in a furnace regulated at 1100.degree. C. Complete melting in 5 hours is followed by pouring. Very slight marbling is observed on the surface, which undoubtedly corresponds to traces of molybdate but is entirely acceptable. Analysis shows that the glass contains 10.2% of Na.sub.2 O for 46% of silica, i.e. a ratio of silica to sodium of 4.5, whereas this ratio is equal to 4.56 in the typical formulation of the final glass. This example demonstrates the possibility of producing, as required, a calcined gel having a composition which is difficult to obtain in the form of a glass frit, and in particular the possibility of producing a low-sodium calcined gel which enables the solution of FP and the soda effluent to be vitrified at the same time. Example 4 This is an attempt to prepare 1 kg of glass immobilizing radioactive waste (solutions of FP), using an inactive matrix of the following composition: ______________________________________ SiO.sub.2 63.4% B.sub.2 O.sub.3 22.7% Na.sub.2 O 11.3% Li.sub.2 O 2.6% ______________________________________ This matrix is prepared by mixing the following solutions in a turbine: This gives a gelled solution which changes to a gel and is dried at 150.degree. C. for 24 h. The solution of FP to be treated in this example is simulated by dissolving the following compounds in 1400 g of water: ______________________________________ Sr(NO.sub.3).sub.2 6.7 g ZrO(NO.sub.3).sub.2.2H.sub.2 O 29.3 g Mn(NO.sub.3).sub.2.4H.sub.2 O 30.3 g Mo 11.3 g Te 1.4 g CsNO.sub.3 13.1 g Ba(NO.sub.3)2 8.7 g Y(NO.sub.3).sub.3.6H.sub.2 O 4.3 g La(NO.sub.3).sub.3.6H.sub.2 O 23.9 g Ce(NO.sub.3).sub.3.6H.sub.2 O 25.1 g Pr(NO.sub.3).sub.3.4H.sub.2 O 12.3 g Nd(NO.sub.3).sub.3.6H.sub.2 O 45.6 g Fe(NO.sub.3).sub.3.9H.sub.2 O 151.8 g Al(NO.sub.3).sub.3.9H.sub.2 O 448.5 g Mg(NO.sub.3).sub.2.6H.sub.2 O 356.1 g Cr(NO.sub.3).sub.3.9H.sub.2 O 21.1 g Ni(NO.sub.3).sub.2.6H.sub.2 O 17.1 g LiNO.sub.3 87.5 g ______________________________________ 240 g of commerical nitric acid (65% by weight) are added to this solution. The solution obtained is stirred for 1 hour, then dried for 24 hours at about 150.degree. C. and then calcined for 4 hours at about 400.degree. C. The resulting calcinate of FP and dried gel are then introduced simultaneously into a crucible. The mixture is melted at 1025.degree. C. for 5 hours. The glass obtained has the following composition: ______________________________________ SiO.sub.2 46% Cs.sub.2 O 0.95% B.sub.2 O.sub.3 16.5% BaO 0.51% Na.sub.2 O 8.2% Y.sub.2 O.sub.3 0.14% Li.sub.2 O 3.8% La.sub.2 O.sub.3 0.90% SrO 0.33% Ce.sub.2 O.sub.3 0.95% ZrO.sub.2 1.35% Pr.sub.6 O.sub.11 0.51% MnO.sub.2 1.05% Nd.sub.2 O.sub.3 1.75% MoO.sub.3 1.7% Fe.sub.2 O.sub.3 3% TeO.sub.2 0.17% Al.sub.2 O.sub.3 6.1% NiO 0.44% MgO 5.6% Cr.sub.2 O.sub.3 0.4% ______________________________________ This glass shows no precipitates or traces of molybdate on the surface. In the tests described, concentrated solutions were prepared (some even being close to saturation point) so as not to increase the drying times or the volumes of liquid to be handled. For reasons of pumping and flows in particular, it may be necessary to dilute these solutions more, but this has no adverse effect on the process. The process developed by the Applicant Company therefore differs from the processes described previously, especially the Westinghouse process. The Applicant Company considers that it has succeeded in preparing, in an aqueous medium, a borosilicate matrix which is ready to be employed for the treatment of nuclear waste, by virtue of the solutions and stirring method used. Stirring at a high rate of shear makes it possible to achieve thixotropic mixing and homogeneity. As soon as stirring stops, the viscosity increases and polymerization rapidly develops, thus "freezing" the ions before they can react (for example precipitation, sedimentation). The process forming the subject of the invention offers an important advantage when operated industrially in a nuclear environment: the matrix is prepared in an inactive environment, so the whole of this part of the process is not subject to the rigid and essential constraints to be observed in an active environment, and the technologies conventionally used in the chemical industry can be employed without modification. Furthermore, the second part of the process (heat treatment with introduction of the waste) can utilize, practically without modification, the existing production lines which are already installed and work with the oxides.
	summary
	abstract	Methods and systems are disclosed for the production of hydrogen and the use of high-temperature heat sources in energy conversion. In one embodiment, a primary loop may include a nuclear reactor utilizing a molten salt or helium as a coolant. The nuclear reactor may provide heat energy to a power generation loop for production of electrical energy. For example, a supercritical carbon dioxide fluid may be heated by the nuclear reactor via the molten salt and then expanded in a turbine to drive a generator. An intermediate heat exchange loop may also be thermally coupled with the primary loop and provide heat energy to one or more hydrogen production facilities. A portion of the hydrogen produced by the hydrogen production facility may be diverted to a combustor to elevate the temperature of water being split into hydrogen and oxygen by the hydrogen production facility.
	description	This application claims the benefit of U.S. Provisional Application No. 63/003,431 filed Apr. 1, 2020, which is incorporated herein by reference in its entirety. The present invention relates generally to systems and vessels for storing and transporting high level radioactive nuclear waste such as used or spent nuclear fuel (SNF), and more particularly to a wet cask storage system with overpressurization protection. A typical module or cask employed to store and transport fissile radioactive waste such as SNF uses an inert gas such as helium to protect the elongated zirconia metal tubes (also referred to as tube cladding) of the fuel rods from oxidation. Such casks insert gas filled casks are referred to as a “dry cask.” Multiple fuel rods are bundled together in a support structure referred to as a fuel assembly which are well known in the art without undue elaboration here. The fuel assemblies are liftable structures typically having a rectangular cuboid shape for U.S. reactors which are configured for insertion into the reactor as self-supporting units. A gas filled dry cask (“dry cask”) holding multiple such fuel assembly units, however, is not perfect from the standpoint of controlling the temperature of the heat-emitting nuclear waste fuel because the heat rejection rate from such a vessel is inhibited by the low thermal conductivity of the gaseous media. This exposes the zirconia fuel rod tubes (cladding) to oxidation and damage, thereby adversely affecting the structural integrity of the containment provided by the tubes for the fissionable nuclear fuel material (e.g. uranium ceramic pellets) packed inside the fuel rods. Improvements in cask storage systems for radioactive nuclear waste is desired. The present application discloses a wet cask system for storing and transporting radioactive nuclear waste such as without limitation spent nuclear fuel (SNF). The system includes an unventilated and hermetically sealed cask containing a volume or inventory of water in which the SNF (e.g. fuel assemblies) is submerged. In one embodiment, the water may be borated for additional radiation shielding. The body of the cask comprises radiation shielding to block and attenuate gamma and neutron radiation emitted by the SNF assemblies. The wet cask system preferably further includes a pressure control sub-system for limiting internal pressure in the cask during a high pressure excursion conditions. In various embodiments, the pressure control sub-system comprises one or more evacuated internal pressure control devices which may be pressure surge capacitors in one embodiment. These capacitors are configured for insertion into the cask cavity occupied by the SNF, and operable to control and mitigate high pressure surge events experienced internally within the wet cask such as those occurring under various postulated accidents and abnormal operating conditions previously described herein. Advantageously, this protects the structural integrity of the cask from such high pressurize excursions which may be caused by external factors (such as fire or degradation of the heat rejection process from the external surface of the cask) or a massive liberation of the gases encapsulated in the fuel rods due to degradation of this metal zirconia cladding described above. Each pressure surge capacitor may be a fully welded and hermetically sealed vessel (no credible path for leakage in or out) in one embodiment comprising at least one rupture disk which seals an internal vacuum chamber inside the capacitor. Each rupture disk is designed and constructed to burst at a predesigned/predetermined pressure level or condition occurring inside the cask cavity external to the pressure surge capacitor. This allows the excess cask pressure occurring during a high pressure excursion to bleed into capacitor, thereby returning the pressure inside the cask to acceptable levels. The vacuum cavity inside each pressure surge capacitor is evacuated to sub-atmospheric (i.e. negative pressure) conditions to the greatest extend practicable. The pressure surge capacitors may have an elongated tubular configuration in some embodiments. The wet cask with hermetically sealable cavity may be used for various applications associated with operation of a nuclear reactor such as in a nuclear power generation facility. For example, in one non-limiting application, the wet cask may be used to transfer spent nuclear fuel assemblies in a continuously submerged stated in the cask between spent fuel pools. The fuel assemblies may be loaded into the wet cask in a first pool underwater, the cask may be lifted out of the first pool and transported to and positioned in a second pool. Radiation blocking is achieved by maintaining the fuel assemblies in the water-impounded cask even during transport. Although the cavity of the cask may be configured and have appurtenances designed to hold SNF assemblies in some embodiments, any type or form of high level radioactive nuclear waste or irradiated materials may be stored in a submerged stated in the inventory of water held by the cask. Such high level radioactive waste materials may be collectively referred to as “radioactive nuclear waste.” In one aspect, a storage system for radioactive nuclear waste comprises: a longitudinal axis; a cask comprising a hermetically sealable internal cavity configured to hold an inventory of water sufficient to submerge the nuclear waste therein; and a pressure surge capacitor disposed in the cask, the pressure surge capacitor comprising a vacuum cavity evacuated to sub-atmospheric conditions; wherein the pressure surge capacitor is configured to suppress a pressure surge in the internal cavity of the cask. In another aspect, a cask with overpressurization protection for storing nuclear waste fuel comprises: a longitudinal axis; a cask body comprising a removable lid assembly, a base, and a circumferential wall including radiation shielding, the cask body forming a hermetically sealed internal cavity configured for holding spent nuclear fuel submerged in an inventory of water; a pressure surge capacitor disposed in the cask, the pressure surge capacitor comprising a vacuum cavity evacuated to sub-atmospheric conditions; and the pressure surge capacitor further comprising at least one rupture disk constructed to burst at a predetermined pressure level inside the cask associated with a cask overpressurization condition; wherein the rupture disk when burst allows a portion of the water to fill the vacuum chamber to reduce pressure inside the cask. In another aspect, a method for controlling pressure in a wet nuclear waste storage system comprises: providing a cask comprising a sealable internal cavity configured for storing nuclear waste; positioning a pressure surge capacitor in the cask, the pressure surge capacitor comprising a vacuum cavity evacuated to sub-atmospheric conditions and in fluid communication with the internal cavity; filling the cask with water; submerging the nuclear waste in the water; and sealing a lid assembly to the cask to hermetically seal the internal cavity; wherein the pressure surge capacitor is configured to suppress a pressure surge in the internal cavity of the cask. The method may further comprise after the sealing step, steps of: increasing the pressure inside the cask to exceed a predetermined burst pressure of a rupture disk of the pressure surge capacitor; bursting the rupture disk; and admitting a portion of the water into the pressure surge capacitor which reduces the pressure inside the cask. All drawings are schematic and not necessarily to scale. Features shown numbered in certain figures which may appear un-numbered in other figures are the same features unless noted otherwise herein. A general reference herein to a figure by a whole number which may include related figures sharing the same whole number but with different alphabetical suffixes shall be construed as a reference to all of those figures unless expressly noted otherwise. The features and benefits of the invention are illustrated and described herein by reference to non-limiting exemplary (“example”) embodiments. This description of exemplary embodiments is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. Accordingly, the disclosure expressly should not be limited to such exemplary embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features. In the description of embodiments disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,”, “above,” “below,” “up,” “down,” “top” and “bottom” as well as derivatives thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a particular orientation. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. As used throughout, any ranges disclosed herein are used as shorthand for describing each and every value that is within the range. Any value within the range can be selected as the terminus of the range. In addition, any references cited herein are hereby incorporated by reference in their entireties. In the event of a conflict in a definition in the present disclosure and that of a cited reference, the present disclosure controls. The terms “seal weld or welding” as may be used herein shall be construed according to its conventional meaning in the art to be a continuous weld which forms a gas-tight hermetically sealed joint between the parts joined by the weld. The term “sealed” as may be used herein shall be construed to mean a gas-tight hermetic seal. FIGS. 1-22 show various aspects and features of the wet cask system for storing and transporting radioactive nuclear waste such as spent nuclear fuel (SNF) according to the present disclosure. The wet cask system advantageously overcomes the shortcomings of insert gas “dry casks” previously described herein. The relatively high conductivity of water keeps the fuel much cooler than does the inert gas medium in the dry cask; an advantage that is highly desirable from the standpoint of maintaining a low pressure inside the nuclear fuel rods. The fuel rods are elongated and thin-walled zirconium metal tubes (also called fuel cladding) containing the fuel pellets of fissionable material (e.g. uranium ceramic pellets) and an inert fill gas mixed with the radioactive gases in the tubes produced in the reactor core. The high-pressure attendant to high temperature (pursuant to the classical Gas law) causes a high membrane stress in the fuel cladding which is known to cause cladding's failure and release of its gaseous contents into the cask's fuel storage cavity. Preventing these harmful gases from escaping into the environment is a key mission of a cask. The cask is therefore designed to withstand a rise in its internal pressure caused by failure of the fuel cladding. Assuming that a significant quantity of gas release could occur during a cask's operation, designing its pressure retention capability with ample structural margin is accordingly a mandatory requirement in virtually all regulatory jurisdictions. In a wet cask in which the spent nuclear fuel (SNF) assemblies are immersed in water, the small free volume above the water and SNF (e.g. headspace) is occupied by water vapor. In case of a regulatory postulated fire event, heating the captive volume of water in the wet cask can raise the vapor pressure in the headspace or an accident leading to massive fuel rod failures inside the cask can release large quantities of the fuel rod gas into the cask cavity. Generation of hydrogen and oxygen by radiolysis of water is another source of pressure build up, although this problem is largely overcome by the use of passively acting hydrogen recombiners or hydrogen “getters” placed inside the cask. The vulnerability of the wet cask to rapid pressure rise is further aggravated by the fact that, at high pressures, even a small increase in the temperature causes a large bump in the saturation pressure. To make wet casks safe and viable for storage/transport of high heat load used or spent nuclear waste fuel, a pressure control sub-system is disclosed herein to protect the cask from a high internal pressure surge under the foregoing accident conditions. With continuing reference in general to FIGS. 1-22, the present wet cask system with integrated pressure control sub-system generally includes a leak-tight sealable cask 100 and at least one pressure surge capacitor 200 operable to absorb a high pressure excursion occurring internally within the cask. There are no provisions for circulating ambient cooling air through cask 100, which is distinguishable from vertical ventilated type overpacks or casks well known in the industry. Cask 100 may be a hermetically sealed, leak-tight pressure vessel comprising a vertically elongated metallic cylindrical body 101 defining a vertical longitudinal axis LA passing through the vertical centerline and geometric center of the body. The cask body generally includes (in progression from top to bottom) lid assembly 110, annular top flange 103, cylindrical circumferential wall 102, and circular base 104 at bottom. Circumferential wall 102 defines a circumferentially-extending sidewall extending vertically between top flange 103 and base 104. The top and bottom ends of wall 102 may be fixedly coupled to the top flange and bottom base via welded connections such as one or more circumferentially continuous seal welds for each to permanently join the components together. Top flange 103 and base 104 may be forged steel structures in one embodiment for added mechanical strength in one embodiment. In some embodiments, the external surface of the cask circumferential wall 102 may optionally comprise a plurality of annular heat transfer fins 118 extending circumferentially around the cask 100. The fins may be arranged in longitudinally spaced apart manner on the cask and extend in a vertical array between top flange 103 and bottom base 104 as shown. Since the sealed cask is not cooled by introducing and flowing ambient cooling air internally through the cask, the fins help remove heat emitted by the decaying fuel in the SNF assemblies in the cask. In other embodiments, the fins may be omitted. The cask body 101 defines an internal cavity 105 which extends longitudinally for a full height of the cask from base 104 at bottom to the top end of circumferential wall 102. The cavity 105 is configured in dimensioned to hold a plurality of spent nuclear fuel (SNF) assemblies 119 (see, e.g. FIG. 10). Cavity 105 is hermetically sealed when lid assembly 110 is removably coupled to the cask body. The fuel assemblies may be insertably contained in a fuel basket 115 is disposed in cavity 105 and seated on the bottom base 104. This design obviates the need for a typical unshielded fuel canister used with some casks. The present cask 100 may instead be completely submerged directly into the spent fuel pool associated with the reactor for loading individual fuel assemblies into the basket while the assemblies remain immersed under water for radiation containment. The fuel basket 115 is a honeycomb prismatic structure comprising an array of vertically-extending openings forming a plurality of vertical longitudinally-extending fuel assembly storage cells 116. Each cell is configured in cross-sectional area and shape to hold a single U.S. style nuclear fuel assembly 119 (see, e.g. FIG. 10) having a rectangular cuboid configuration, which in turn contains a multitude of spent nuclear fuel rods 119a previously described herein (or other radioactive nuclear waste). The cells 116 may each have generally square cross-sectional shape as shown which is complementary configured to the cross sectional shape of the fuel assembly. Such fuel assemblies and the foregoing fuel basket structure are well known in the industry. The fuel basket may be formed in various embodiments by a plurality of interlocked and orthogonally arranged slotted plates built up to a selected height in vertically stacked tiers of plates. Examples of slotted plate basket constructions are disclosed in commonly-owned U.S. patent application Ser. No. 17/115,005, which is incorporated herein by reference. Other constructions of fuel baskets such as multiple laterally adjacent vertically extending tubes or other structures to the canister baseplate may be used and others used in the art may be used. In addition, the fuel assembly cells 116 in some constructions may have a hexagonal cross-sectional shape to accommodate hexagonal fuel assemblies commonly used in Russia. The fuel basket construction however is not limiting of the present invention. As best shown in FIG. 19, the polygonal shaped structure of the complete fuel basket 115 structure fitted inside the cylindrical internal cavity 105 of cask 100 leaves a plurality of unused peripheral areas or regions 117 between the cask circumferential wall 102 and basket. These regions have a par-polygonal shape comprising one outer non-polygonal side formed by an arcuate portion of the cask wall 102 and remaining polygonal inner sides of linear shape formed by parts of the fuel basket 115. Peripheral regions 117 are dead zones serving typically no function and considered wasted space. Accordingly, such peripheral regions are generally kept to a minimum as much as practical. The present cask pressure control sub-system however advantageously makes use of these dead zones, as further described herein. Referring initially in general to FIGS. 1-14 and 20-22, wet cask 100 is a heavy radiation shielded nuclear waste fuel storage and transport vessel having a composite wall construction operable to ameliorate the gamma and neutron radiation emitted by the SNF fuel assemblies contained therein to safe levels outside the cask. Circumferential wall 102 of the cask comprises in progression from inside to outside inner shell 120 adjacent to cask cavity 105, intermediate shell 121, and outer shell 122 (see, e.g. FIGS. 11, 12, and 21). Inner shell 120 may be formed of thick steel. Shells 120-122 are coaxially aligned around longitudinal axis LA and radially spaced apart to permit radiation shielding materials to be located in the annular gaps or spaces formed between the shells. Gamma shielding material 123 is disposed in inner annular space 125 between inner shell 120 and intermediate shell 121. Any suitable gamma shielding material may be used, including lead as shown, concrete, or others. In one embodiment, gamma shielding material 123 may be provided in the form of longitudinally-elongated arcuately curved blocks each extending for a majority of the height of the cask 100 at least covering a portion of the height of the fuel basket which contains the SNF fuel assemblies 119. A plurality of such blocks are arranged circumferentially around the cask encircling internal cavity 105. The gamma shielding blocks may be separated by conductive inner radial ribs 127 welded between and to inner shell 120 and intermediate shell 121 (see, e.g. FIG. 21). Ribs 127 may be made of steel in one embodiment. A plurality or array of circumferentially spaced apart ribs 127 encircle the inner shell 120. The ribs 127 form longitudinally-extending pockets which receive and organize the blocks of the gamma shielding material. Notably, ribs 127 further act as thermally conductive elements which draw the heat emitted by the SNF assemblies 119 outwards towards the outer shell 122 and heat transfer fins 118 since there is no ambient ventilation air circulated through this unventilated cask. Ribs 127 further provide structural reinforcement for the cask and maintain the annular space 125 between the shells 120, 121. Boron-containing neutron shielding material 124 is disposed in outer annular space 126 between outer shell 122 and intermediate shell 121. Any suitable neutron shielding material containing boron may be used, such as for example without limitation Holtite™ from Holtec International of Camden, Jersey. Other boron-containing materials however may be used. In one embodiment, neutron shielding material 124 may be provided in the form of longitudinally-elongated bars each extending for a majority of the height of the cask 100 at least covering a portion of the height of the fuel basket which contains the SNF fuel assemblies. A plurality of such bars are arranged circumferentially around the cask encircling internal cavity 105 which holds the SNF fuel assemblies 119. The neutron shielding bars may be similarly separated by an outer second plurality or array of conductive outer radial ribs 128 welded between and to outer shell 122 and intermediate shell 121 (see, e.g. FIG. 21). The ribs form longitudinally-extending pockets which receive and organize the bars of the neutron shielding material (e.g. Holtite™), as well as providing the same heat transfer function and structural reinforcement for the cask as ribs 127 described above in addition to maintaining the annular space 126 between the shells 121, 122. Outer radial ribs 128 may be formed of copper in one embodiment to maximize heat transfer between the intermediate shell 121 and outer shell 122. In some embodiments, the forgoing inner radial ribs 127 and/or the outer radial ribs 128 may each be formed as integral parts of an annular or ring-shaped monolithic casting. Each of the castings may then be fitted between the shells 120-122 in the circumferential wall 102 of the cask 100 in their respective positions described above. In one embodiment, the outer radial rib 128 casting may be made of copper to maximize heat transfer. Inner radial rib 127 casting may be formed of steel in some embodiments if used. Alternatively, either of the inner or outer radial ribs 127, 128 may be welded directly to their respective shells which they bridge. Lid assembly 110 may comprise a lower inner lid 111 and upper outer lid 112 stacked thereon (best shown in FIGS. 3-4 and 12). Inner lid 111 is configured for partial insertion into cask cavity 105 and terminates proximate to the top of the fuel basket 115. A lower portion of the lid 111 therefore has a smaller diameter than the inside diameter of the cask defined by inner shell 120. An upper portion of inner lid 111 has a larger diameter radially protruding annular flange 111a seated on a mating step-shaped annular shoulder 113a on the top of the cask defined by cask top flange 103. Lid 111 may include a centrally located lifting lug 113 on top configured to be grasped by a grappling assembly of a hoist or crane for lifting the lid into place on the cask 100. Lifting lug 113 may be disk-shaped in one embodiment. Lug 113 is received and nested in a downwardly open complementary configured circular recess 114 formed on the bottom of the outer lid 112. Outer lid 112 has a larger diameter than the inner lid and comprises a radially protruding annular flange 112a seated on a mating step-shaped annular shoulder 114a on the top of the cask also defined by cask top flange 103. Outer lid 112 is bolted to top flange 103 of cask 100 by a circular array of closure bolts 112b, thereby trapping the inner lid 111 onto the cask. Lid assembly 110 further comprises a plurality of annular seals 150 compressed between the cask body (e.g. top flange 103) and each of the inner and outer lids 111, 112 (best shown in FIG. 12). When lid assembly 110 is coupled to cask 100 (e.g. bolted), a hermetically sealed leak-tight cask internal cavity 105 and pressure vessel is created which is fluidly isolated from the ambient environment. Cask 100 may further include a plurality of radially protruding lifting lugs 130 for maneuvering the cask such as lifting into and out of the spent fuel pool during the process of loading SNF assemblies 119 into the cask fuel basket 115. At least one bottom drain assembly 160 may be provided which is openable/sealable to drain the inventory of water in the cask in which the fuel assemblies are submerged (FIG. 11). Drain assembly 160 may be formed in base 104 in some embodiments. A top openable/sealable port 161 which is fluidly coupled to cask internal cavity 105 via a duct as shown in FIG. 12. Port 161 may be used for various purposes, including for example without limitation for testing the conditions inside the cask, or optionally to convert the cask 100 to long dry storage by circulating an inert gas (e.g. helium) through the cask to dry cavity 105 in conjunction with the bottom drain assembly 160 for establishing a gas flow path therethrough. Inert gas cask drying systems are well known in the art without further elaboration. Top port 161 may be a gas inlet and bottom drain assembly 160 may be a gas outlet, or vice versa. Top port 161 may be formed in top flange 103 in some embodiments. As previously described herein, wet cask 100 is a water-impounded cask in which the fuel assemblies 119 are immersed under water. The water W has a surface level sufficient to at least fully cover the fuel assemblies. An exemplary surface level of water W is represented in FIG. 12 by the dashed line. The pressure control sub-system comprising pressure surge capacitor 200 operable to absorb a high pressure excursion occurring internally within the cask will now be further described. FIGS. 13-17 show the pressure surge capacitor in isolation. Referring initially to FIGS. 13-17 and 21-22, pressure surge capacitor 200 has a longitudinally elongated cylindrical tubular body defining a vertical centerline and comprising a top end 201, bottom end 202, and cylindrical sidewall shell 203 extending therebetween and defining an internal pressurizable vacuum space or chamber 204 having a volume V1. The terminal end portions of the capacitor 200 define end caps 206 having a thickness measured parallel to the vertical centerline Vc which is substantially greater (e.g. 3 times or more) than the wall thickness of the sidewall shell 203 (measured transversely to centerline Vc). As shown in FIGS. 11 and 22, the pressure surge capacitor 200 is a longitudinally elongated pressure vessel having a greater longitudinally length than its diameter. Capacitor may have a height at least coextensive with the height of the fuel basket 115 in some embodiments. Capacitor 200 therefore has a height which extends for a substantial majority of the height of the internal cavity 105 of cask 100 from proximate to the bottom of lid assembly 110 to base 104 of the cask. In one embodiment, pressure surge capacitor 200 is positioned and located adjacent to inner shell 120 of the cask in its internal cavity 105, such as in one of the larger peripheral regions 117 inside cask 100 lying between the fuel basket 115 and inner shell 120 (see also FIG. 19). This otherwise dead space too small to accommodate a full rectangular SNF fuel assembly is advantageously not wasted and advantageously used in a cask overpressurization function. Although one pressure surge capacitor 200 is shown, other embodiments may place multiple capacitors 200 in a similar manner in peripheral regions 117 for added cask pressure surge protection as needed. A flow inlet opening 205 is formed through at least one end 201 or 202 of pressure surge capacitor 200 (e.g. end caps 206), and in some embodiments through both ends at top and bottom as shown. Inlet opening 205 is in fluid communication with vacuum chamber 204 of pressure surge capacitor 200 for selectively admitting high pressure water held inside cask cavity 105 during a cask n internal pressure excursion (increase). In one embodiment, inlet opening 205 may be circular in transverse cross section and comprises a larger diameter outer portion 209 and small diameter inner portion 210. A step-shaped annular shoulder 211 is formed therebetween (best shown in FIGS. 16 and 17). Each inlet opening 205 is fitted with a pressure relief device 220 comprising a circular metallic rupture disk 221 and annular disk retaining ring 222. Retainer ring 222 includes a central opening 223 which allows pressurized water to flow through the inlet opening 205 into the pressurizable vacuum chamber 204 of the tubular pressure surge capacitor when the rupture disk bursts. Rupture disk 221 is designed and constructed with predetermined burst pressure selected to protect the cask 100 and fuel assemblies 119 therein from a potentially damaging high pressure condition previously described herein internal to the cask caused by degradation and/or failure of the fuel rod cladding. The predetermined burst level is set taking into consideration the differential between the sub-atmospheric vacuum condition inside the pressure surge capacitor 200 and the pressure outside the capacitor inside the cask cavity 105. Any suitable type of metal rupture disk may be used, including without limitation a reverse buckling design as shown herein (in which the convex side of the rupture disk faces the high pressure source) or a forward-acting disk design (in which the concave side of the disks faces the high pressure source). During assembly of each pressure relief device 220, one rupture disk 221 is positioned and seated on an outward facing disk seating surface 224 formed at the innermost end of outer portion 209 of flow inlet opening 205. One retaining ring 222 is then positioned over the rupture disk and coupled to the end cap 206 of pressure surge capacitor 200 such as via welding, threaded connection, or other. This traps the rupture disk 221 between the retaining ring and seating surface 224. The dome shaped central portion of the rupture disk protrudes outwards into and partially enters the central opening 223 of retaining ring 222 where it is exposed to the internal pressure of cask 100 inside cavity 105. Each pressure surge capacitor 200 is then evacuated to as deep a vacuum (negative pressure) as practicable. A vacuum port 230 may be formed in sidewall shell (or alternatively the end caps 206) for evacuating the vacuum chamber 204 of the capacitor. A valve 231 may be removably coupled to the port 230 for drawing the vacuum via an external vacuum pump 232 (valve and pump shown schematically in FIG. 18). Any suitable type of valve may be used which is configured with a suitable end fitting configuration for detachable coupling to a hose or other flow conduit line fluidly connected to the vacuum pump. The evacuated pressure surge capacitor(s) 200 are now ready for deployment and operation. Each pressure surge capacitor 200 provided (e.g. one or more) may be positioned inside cask cavity 105 in an available empty space such as open peripheral regions 117 (see, e.g. FIG. 19). Capacitors 200 may be loosely positioned in the cask, or alternatively may be fixedly attached to the outside walls of the fuel basket 115 (such as via welding) before the fuel basket is installed in cask 100. In the latter case, the capacitors may be evacuated before or after welding to the fuel basket. Capacitors 200 are constructed via selection of the type of metal used for the body and end caps (i.e. mechanical strength and other material properties), and associated thicknesses to withstand the external pressure which the cask cavity will exert from the elevated hydraulic pressure that would result under a condition of elevated temperature of the body of water inside cask 100. The pressure surge capacitors are fully exposed to the temperature and pressure conditions inside the cask cavity 105. In operation, if an overpressurization condition should occur in cask cavity 105 which exceeds the pre-designed and predetermined burst pressure of the rupture disk 221, the disk will burst allowing the excess pressure to bleed into the evacuated vacuum chamber 204 of the capacitors 200 (see dashed water inflow arrows in FIG. 18). The internal cask pressure will attempt to equilibrate inside and outside the pressure surge capacitors in the cask cavity 105 to thereby lower the internal cask pressure to a stable acceptable pressure level, thereby ameliorating the high pressure excursion condition. Although only a single pressure surge capacitor 200 is shown for clarity of depiction in the figures, it will be appreciated that other embodiments will include any suitable number of pressure surge capacitors as needed to provide the surge capacity necessary to compensate for and ameliorate the postulated cask overpressurization conditions that could possibly occur during storage of the SNF in sealed wet cask 100. A method for controlling pressure in a sealed cask 100 using pressure surge capacitors 200 will now be briefly summarized. The method generally comprises providing an unventilated cask 100 comprising a sealable internal cavity 105 configured for storing nuclear waste such as spent nuclear fuel assemblies 119; positioning a pressure surge capacitor 200 in the cask, the pressure surge capacitor comprising a vacuum cavity 204 evacuated to sub-atmospheric conditions and in fluid communication with the internal cavity; filling the cask with water; submerging the nuclear waste in the water; and sealing a lid assembly 110 to the cask to hermetically seal the internal cavity; wherein the pressure surge capacitor is configured to suppress a pressure surge in the internal cavity of the cask. The method may further include after the sealing step, steps of: increasing the pressure inside the cask to exceed a predetermined burst pressure of a rupture disk 221 of the pressure surge capacitor 200; and admitting a portion of the water into the pressure surge capacitor which reduces the pressure inside the cask. The pressure surge capacitor therefore advantageously operates to relieve the cask pressure and ameliorate the high pressure increase excursion. In some embodiments, the filling step includes lowering the cask 100 into a first spent fuel pool 250 below a water surface S thereof (schematically shown in FIG. 11). The method may further include after the sealing step, steps of: lifting the cask out of the first spent fuel pool; and transporting the cask to a second spent fuel pool. The cask may be lowered into the second fuel pool for either loading additional SNF assemblies 119 into the cask, or unloading the spent fuel assemblies into the second fuel pool such as into cavities of a SNF storage rack such as those disclosed in commonly-owned U.S. Pat. No. 10,847,274, which is incorporated herein by reference in its entirety. Variations in the foregoing steps of the method, and additional steps, may be used. It bears noting pressure surge capacitor 200 is shown having a cylindrical configuration, in other embodiments the capacitor may have a body shaped other than cylindrical with circular transverse cross section, such as any suitable non-polygonal or polygonal configuration. The shape of the pressure surge capacitor does not limit the invention. FIG. 21 shows an alternative embodiment of the pressure control sub-system of the cask 100 in which one or more pressure surge capacitors 200a are incorporated and embedded in the circumferential walls 102 of the main cylindrical body of the cask. Inner shell 120 includes one or more separate flow apertures 151 fluidly coupled between each rupture disk 221 provided for the capacitor 200a and the internal cavity 105 of the cask 100. The vacuum chamber 204 of the capacitor is therefore fluidly connected to the cavity 105 through the rupture disk(s) 221 previously described herein and function in the same manner to protect the cask from internal high pressure surges/excursions. A plurality of embedded pressure surge capacitors 200a may be provided. FIG. 22 shows a second alternative embodiment of the pressure control sub-system of the cask 100 in which one or more pressure surge capacitors 200b are incorporated in the lid assembly 110 of the cask 100. In this illustrated embodiment, a centrally located pressure surge capacitor 200b is fixedly coupled to the bottom surface of inner lid 111 (such as via welding). This locates capacitor 200b in the headspace between the bottom surface of inner lid 111 and the top edges of the fuel basket within the internal cavity 105 of the cask 100. Capacitor 200b may have a cylindrical body similar to capacitor 200 previously described herein and includes at least one pressure relief device 220 (i.e. rupture disk 221 and retaining ring 222) in a bottom surface of the lid-mounted capacitor 200b. In this embodiment, the diameter of capacitor 200b may be larger than its longitudinal height. In some embodiments, an array comprised of multiple lid-mounted pressure surge capacitors 200b may instead be provided. It bears noting that alternative pressure surge capacitors 200a and 200b may be provided instead of pressure surge capacitors 200 previously described herein which are located directly in the cask fuel storage internal cavity 105, or alternatively in addition thereto if added pressure surge amelioration capacity is needed. While the foregoing description and drawings represent some example systems, it will be understood that various additions, modifications and substitutions may be made therein without departing from the spirit and scope and range of equivalents of the accompanying claims. In particular, it will be clear to those skilled in the art that the present invention may be embodied in other forms, structures, arrangements, proportions, sizes, and with other elements, materials, and components, without departing from the spirit or essential characteristics thereof. In addition, numerous variations in the methods/processes described herein may be made. One skilled in the art will further appreciate that the invention may be used with many modifications of structure, arrangement, proportions, sizes, materials, and components and otherwise, used in the practice of the invention, which are particularly adapted to specific environments and operative requirements without departing from the principles of the present invention. The presently disclosed embodiments are therefore to be considered in all respects as illustrative and not restrictive, the scope of the invention being defined by the appended claims and equivalents thereof, and not limited to the foregoing description or embodiments. Rather, the appended claims should be construed broadly, to include other variants and embodiments of the invention, which may be made by those skilled in the art without departing from the scope and range of equivalents of the invention.
	summary
	summary
	description	The information setting forth the placement of fuel bundles, each of which has various attributes, in a nuclear reactor core is referred to as the loading map. In conventional core design, creating the loading map is an experienced based, trial and error, iterative process. The core designer generally receives plant specific critical to quality factors such as plant cycle energy requirements, thermal and operational limits, shut down margins, etc. The core designer will also have information on the layout of the reactor core; namely, an indication of the how the nuclear fuel bundles are positioned within the core. Some of the critical to quality factors may even concern the layout. For example, the core designer may receive input requiring the positioning of certain fuel bundles within the layout. Given this information, the core designer then makes a guess, based on experience and various rules of thumb he may have developed over time, on the initial positioning of fuel bundles in the reactor core. Specifically, the core designer guesses how many fresh fuel bundles to place in the core, and what types of fresh fuel bundles to use. A fresh fuel bundle is a fuel bundle that has not been exposed. Fuel bundles of the same type have substantially the same attributes. The attributes include but are not limited to: uranium loading, average enrichment, gadolinia loading, number of axial zones, product line, and thermal-mechanical characteristics of the fuel bundles. Different types of fresh fuel bundles have one or more different attributes. In deciding how many fresh fuel bundles to use, the core designer is also deciding how many of the fuel bundles currently in the core to reuse. Reusing the fuel bundles currently present in the core can mean leaving a fuel bundle in its existing location, or moving the fuel bundle to a different location in the core. As part of the core design, the core designer also determines other operational parameters of the reactor core such as control blade positions, core flow, etc. Having specified these operational control parameters, a Nuclear Regulatory Commission (NRC) licensed simulation program is then run on the initial core design. Based on the results of the simulation, the core designer utilizes experience and rules of thumb to fix perceived problems in the design and, in general, improve the design; particularly with respect to the critical to quality factors. These changes may include changing the loading map. The process repeats until the core designer is satisfied with the design. In an embodiment of the present invention, at least one fuel bundle within a core is assigned into one of a plurality of bundle groups, the assignment being based on an associated fuel bundle characteristic. For example, the fuel bundle characteristic may be an exposure characteristic. In another embodiment of the present invention, one of a plurality of bundle groups is selected. A grouping operation is performed on the selected bundle group. In another example embodiment of the present invention, a plurality of fuel bundle groups in a current loading map is created. A first plurality of proposed loading maps is generated to establish potential core loading at a next cycle of operation. The plurality of proposed loading maps are filtered to create a second plurality of proposed loading maps, the second plurality of proposed loading maps satisfying at least one fuel bundle group constraint. An objective function is configured to evaluate the second plurality of proposed loading maps. To better understand the present invention, an example method of creating a template and loading map will be described. Then, a process of generating bundle groups within a loading map will be described. Finally, example implementations of bundle groups being used as a constraint during an optimization process will be given. FIG. 1 illustrates an embodiment of an architecture according to the present invention. As shown, a server 10 includes a graphical user interface 12 connected to a processor 14. The processor 14 is connected to a memory 16. The server 10 is directly accessible by a user input device 18 (e.g., a display, keyboard and mouse). The server 10 is also accessible by computers 22 and 26 over an intranet 20 and the Internet 24, respectively. The operation of the architecture shown in FIG. 1 will be discussed in detail below. Creating a Template A user via input 18, computer 26 or computer 22 accesses the server 10 over the graphical user interface 12, and runs a loading map editor program stored in memory 16 according to an exemplary embodiment of the present invention. The loading map editor provides for creating and editing a graphical representation of a nuclear reactor core referred to as a template. However, another form of conveying this information, such as a text file, may also be thought of as the template. FIG. 2 illustrates a quarter-core screen shot of a partially completed template designed according to the methodologies of the present invention using the loading map editor of the present invention. When the loading map editor is initially run, the user has the option via a file menu 30 to access a previously created template or to begin a new template. Assuming the user begins a new template, the loading map editor requests the user to identify the nuclear reactor for which the template is being created. The loading map editor then retrieves the geometry of the identified nuclear reactor from a relational database containing nuclear reactor plant characteristics stored in the memory 18. The loading map editor then displays a blank colorless fuel bundle field 36 of the appropriate size based on the retrieved plant characteristics with the rows and columns numbered (such as with the fuel bundle position Row 6, Column 3 in FIG. 2). Within the fuel bundle field 36, the user may then, for example, using a mouse associated with the input 18, computer 26 or computer 22 click on the fuel bundle positions 38 in the array of possible fuel bundle positions to identify the type (fresh, reinsert, or locked) and setting of the actual fuel bundle in that position. The user may then assign fuel bundles to one of a plurality of fuel bundle sets. Examples of criteria upon which a user may select the bundle set assignment will be described later. As shown on the right side of FIG. 2, the loading map editor provides several tools for performing this assignment task. Specifically, the tools include the headings Load Type 40, Bundle Setting 50 and Numbering Mode 60. An example bundle set selection will now be described. Under the Load Type 40 tool heading, the loading map editor includes a Fresh radio button 42, a Reinsert radio button 44 and a Locked radio button 46. The Fresh, Reinsert and Locked radio buttons 42, 44 and 46 correspond to fresh, reinsert and locked fuel bundle categories. The user, for example, clicks on the desired radio button to choose the desired category and then clicks on the fuel bundle position 38 in the fuel bundle field 36 to assign that category to the fuel bundle position 38. The fresh fuel bundle category indicates to insert fuel bundles that have not been exposed. The loading map editor then displays “F” and a number “N” at the bottom of the fuel bundle position 38. The “F” indicates the fresh fuel bundle category, and the number “N” indicates the Nth fresh bundle type 38. As will be appreciated, the loading map editor maintains a count of the number of fuel bundle types assigned to the core. Multiple bundle positions can be assigned the same bundle type by specifying the same “F” and “N” value for each position. The locked fuel bundle category indicates that a fuel bundle currently occupying an associated fuel bundle position in an actual nuclear reactor core is to remain in that position in creating a new nuclear reactor core loading map. The loading map editor displays “L” and a number “N” in the fuel bundle position 38 when the locked fuel bundle category is assigned. The “L” indicates the locked fuel bundle category, and the number “N” indicates the Nth locked bundle set. The reinsert fuel bundle category indicates to insert a fuel bundle that has been exposed. The loading map editor displays only a number “N” in the fuel bundle position 38 when the reinsert fuel bundle category is assigned. The number indicates a priority of the fuel bundle position 38. The number and the priority indicated by the number will be described in detail below with respect to the Numbering Mode 60 heading. In an exemplary embodiment, the loading map editor displays the fuel bundle positions 38 in a color associated with the assigned category. For example, fresh are displayed in blue, locked are displayed in yellow, and reinserted are displayed in violet. Under the Bundle Setting 50 heading, the loading map editor includes a “1” radio button, a “2” radio button, a “4” radio button, and an “8” radio button. When the “1” radio button is selected by the user, for example, by clicking on the “1” radio button, the category assigned by the user to a fuel bundle position 38 is associated only with the fuel bundle position 38 chosen. Selecting the “2” radio button and assigning a category to a fuel bundle position 38 causes the category to be assigned to the selected fuel bundle position as well as the fuel bundle position 180 degrees symmetric to the selected fuel bundle position. Selecting the “4” radio button causes the loading map editor to request the user to choose between rotational and mirror symmetry. Rotational symmetry is an image property indicating there is a center point around which the object is turned a certain number of degrees and the object still looks the same (i.e., it matches itself a number of times while it is being rotated,). Mirror symmetry (or line symmetry) indicates a correspondence in size, shape, and relative position of parts on opposite sides of a dividing line. If the user assigns a category to a fuel bundle position when rotational symmetry is chosen, this causes the category to be assigned to the selected fuel bundle position as well as the fuel bundle position 38 in each of the other quadrants rotationally symmetric to the selected fuel bundle position. If the user assigns a category to a fuel bundle position when mirror symmetry is chosen, this causes the category to be assigned to the selected fuel bundle position as well as the fuel bundle position in each of the other quadrants symmetric to the selected fuel bundle position. Selecting the “8” radio button causes the loading map editor to consider the total fuel bundle field 36 as octant symmetric—eight symmetric pie pieces. Assigning a category to a fuel bundle position when the “8” radio button is selected causes the category to be assigned to the selected fuel bundle position 38 as well as the fuel bundle positions 38 in each of the other eight pie pieces symmetric to the selected fuel bundle position 38. Under the Numbering Mode 60 heading, the loading map editor includes an Automatic radio button 62 and a Manual radio button 64. Choosing between an automatic numbering mode by selecting the Automatic radio button 62 and a manual numbering mode by selecting the Manual radio button 64 is only permitted when the Reinsert radio button 44 or Fresh radio button 42 has been selected. The numbering mode in general is inapplicable when the Locked radio button 46 is selected. When the Automatic radio button 62 is selected, the loading map editor, which maintains a count of the number of fuel bundle positions 38 assigned the reinsert fuel bundle category, assigns the count plus one to the next fuel bundle position 38 assigned the reinsert fuel bundle category. The assigned number is displayed at the bottom of the fuel bundle position 38. Likewise, the loading map editor maintains a count of the fresh bundle types. When a fuel bundle position 38 is assigned the fresh bundle category the count plus one, referred to above as N, is assigned to that position. “F” and the value of N are displayed at the bottom of the fresh fuel bundle position. When the Manual radio button 64 is selected, the loading map editor maintains the count of the number of fuel bundle positions 38 assigned the reinsert fuel bundle category, but does not assign numbers to the fuel bundle positions 38. Instead, the user may position a cursor in the fuel bundle position 38 and enter the number manually. As alluded to above, the assigned numbers represent assigned priorities. The priorities indicate an order for loading exposed fuel bundles based on an attribute of the exposed fuel bundles. The attributes include, but are not limited to, K infinity (which is a well-known measure of the energy content of the fuel bundle), exposure of the bundle (which is accumulated mega-watt days per metric ton of uranium in the bundle), residence time of the bundle (which is how long the bundle has been resident in the nuclear reactor core), etc. In one exemplary embodiment, the shade of the color associated with the reinserted fuel bundle positions varies (lighter or darker) in association with the assigned priority. The loading map editor according to the present invention also provides several viewing options via a view menu 34 and a zoom slide button 70. Adjusting the zoom slide button 70 by clicking and dragging the zoom slide button 70 to the left and the right decreases and increases the size of the displayed fuel bundle field 36. Under the view menu 34, the user has the option to view a single quadrant of the template, or a full core view of the template. Additionally, the user can control whether certain template attributes are displayed. Specifically, the view menu 34 includes the options of displaying the following in the loading template: control blades, bundle coordinates, core coordinates, etc. Having created the loading template, the user may save the template, or even a partially created template, to the memory 18 by selecting either the “Save” or “Save As” option in the file menu 30. As discussed above, instead of creating a new template, a previously created template may be viewed and, optionally, edited. Using the file menu 30, the user selects an “open” option. The loading map editor then displays the accessible templates stored in the memory 18 or a directory of memory 18. The user then selects an accessible template, for example, by clicking on one of the accessible templates. The loading map editor will then display the chosen template. The user may then edit the chosen template. For example, after selecting a fuel bundle position 38 the user may select under the edit menu to “clear” the category assigned to the fuel bundle position 38. Besides the category assigned to this fuel bundle position 38, the loading map editor also clears the category assigned to associated fuel bundle positions 38. Associated fuel bundle positions 38 are those fuel bundle positions 38 that were assigned the fuel bundle category along with the fuel bundle position 38 selected for clearing because of the bundle setting chosen when the category was assigned to the fuel bundle position 38 chosen for clearing. When fuel bundle positions 38 assigned the fresh or reinserted category are cleared, the loading map editor adjusts the numbering associated with that category. In the case of the fresh bundle category, this is a conditional action based on whether other bundle positions have been assigned the same fresh bundle type. Specifically, the loading map editor performs a cascade operation such that fuel bundle positions assigned the same category and having higher numbers are renumbered in sequence beginning from the lowest number of a deleted fuel bundle position. For example, if reinsert bundle positions numbered 44, 43 and 42 were cleared, then reinsert bundle position having number 45 would be renumbered 42, reinsert bundle position having number 46 would be renumbered 43, etc. The loading map editor also changes the total count of fuel bundle positions assigned the category being cleared. When unassigned bundle positions are created through editing, the user may then newly assign categories to the unassigned bundle positions in the same manner and using the same tools to create a template as described above. In so doing, the user may decide to manually assign, for example, an existing priority to a newly assigned reinsert fuel bundle position. In this instance, the reinsert fuel bundle position already having this number and each reinsert fuel bundle position having a higher number are incremented by one. As a further alternative, the user may want to adapt an existing template for one reactor to another reactor of the same size and physical bundle configuration. To do this, the user may use the “save as” feature in the file menu 30 to create a duplicate of the loading template. Subsequent changes to the bundle field will then apply to the copied template. In addition to creating a template from ‘scratch’ or editing an existing template, the user may have the loading map editor derive a template from a previously loaded core. In the loading map editor, using the file menu 30, the user selects an “auto-generate template” option. The loading map editor then displays a list of the accessible fuel cycles stored in the memory 18. Each fuel cycle corresponds to an actual loading map for a fuel cycle of a nuclear reactor. As will be appreciated, the memory 18 may store loading maps for cycles of different nuclear reactors. Accordingly, the list of cycles displayed by the loading map editor identifies both the nuclear reactor and the cycle. From the list the user selects the cycle (hereinafter “the selected cycle”) that the template will be derived from. The loading map editor then accesses the loading map for the selected cycle. The user is then presented with a dialog box for entering input parameters of the derivation process. The input parameters include: a primary attribute (e.g., exposure, K infinity, etc.) for deriving the template, a tolerance level (discussed in detail below), set list members (8, 4, or 2 bundle settings), bundle symmetry for sets of 4, and a maximum number of assignments to each set list member. For example the user may enter K infinity as the primary attribute, and a tolerance level of 0.2 (which, as described in detail below, is used for forming bundle sets). The user may further enter that sets of 8 and 4 are permitted, the sets of 4 should have mirror symmetry and that a maximum of 14 sets of 4 are permitted. In an exemplary embodiment, the loading map editor provides the user with a drop-down menu. The user selects list members desired for the template from the options given in the drop-down menu. These options include: sets of 8, 4 and 2; sets of 8 and 4; sets of 8 (which forces sets of 4 on the minor axis of the reactor core template); and sets of 4 and 2. In selecting the maximum number of assignments for each set, the user enters this data in the order of the smallest to the largest set size. However, the maximum number of assignments for the largest sets is not entered by the user, as this value is automatically determined based on the maximum number of assignments for the smaller sets. Once the user enters the input parameters, the loading map editor will begin generating a template. First the loading map editor asks the user if locked bundle positions are permitted, if so, then the loading map editor requests the user to identify the cycle previous to the selected cycle in the same manner that the selected cycle was identified. The loading map editor then compares the loading map for the selected cycle with the loading map for the previous cycle of the identified nuclear reactor. Specifically, for each bundle position in the reactor, the loading map editor determines if loading maps for the selected and previous cycles have a bundle with the same serial number in the same bundle position. If so, the bundle position is assigned the locked fuel bundle category in the loading template. After the locked fuel bundle positions are identified, the loading map editor identifies the fresh fuel bundle positions. Specifically, for each bundle position not already identified as a locked bundle position, the loading map editor determines from the characteristics of the selected loading map if the fuel bundle in that bundle position is a fresh fuel bundle. For each identified fresh fuel bundle, the loading map editor also determines the type of fresh fuel bundle from the characteristics of the selected loading map. The loading map editor then assigns the fresh fuel category to the associated fuel bundle position in the template and assigns a type count number N to the fuel bundle position. For each type of fresh fuel bundle located in the selected loading map, the loading map editor assigns a count value to that type. This count value is then assigned to the bundle position along with the fresh fuel bundle category assignment so that fresh fuel bundle positions that should have the same type of fresh fuel bundle are identified by the same value ‘N’ in the loading template. Next, the loading map editor determines whether the identified fresh bundle category positions form any bundle sets. As discussed above, the user identifies the bundle set members permitted in the template. The bundle set members form a set members list. For each bundle position assigned the fresh fuel bundle category, the loading map editor first determines if the bundle position (hereinafter the “current bundle position”) has already been assigned to a set. If so, then the loading map editor proceeds to the next bundle position. If not, then the loading map editor selects the largest set from the set member list and identifies each of the bundle positions that form such a set with the current bundle position. If each of the bundles positions forming the set has been assigned the fresh bundle category and are of the same type as the current bundle position, then the loading map editor records the set of bundle positions as a set. If each of the bundle positions forming the set has not been assigned the fresh bundle category or one of the bundles is not the same type as the current bundle position, then the loading map editor performs the above-described process for the next largest bundle set in the set member list. This process keeps repeating until a set is formed or there are no more sets in the set member list to test. If the members of the set member list have been tested, and no set has been formed, then the current bundle position is recorded as not belonging to a set. Next, the loading map editor identifies the reinserted fuel bundle positions. The bundle positions of the template not assigned to the locked or fresh fuel bundle categories are assigned the reinserted fuel bundle category. Then, the loading map editor determines whether the reinserted bundle category positions form any bundle sets. For each bundle position assigned the reinserted fuel bundle category, the loading map editor first determines if the bundle position (hereinafter the “current bundle position”) has already been assigned to a set. If so, then the loading map editor proceeds to the next bundle position. If not, then the loading map editor selects the largest set from the set member list and identifies each of the bundle positions that form the set with the current bundle position. If each of the bundles positions forming the set has not been assigned the reinserted bundle category, then the loading map editor determines if the next largest set in the set member list includes all reinserted fuel bundle positions. If no set from the set member list results in a set of reinserted fuel bundles, then the loading map editor records the current fuel bundle position as not belonging to a set. Once a set has been formed, the loading map editor calculates the average attribute value for the set. As discussed above, the user identified a primary attribute to use in deriving the template. Here, the loading map editor uses that attribute value for each fuel bundle in the selected loading map forming the associated set in the template to calculate the average attribute value. The loading map editor then determines if the attribute value for each fuel bundle in the set is with the tolerance level from the average attribute. Again, here, the tolerance level was a user input design parameter as discussed above. If the attribute value for each fuel bundle in the set is within the tolerance level of the average attribute value, then the loading map editor records the associated fuel bundle positions in the template as belonging to a set. Otherwise, the loading map editor performs the above-described process for the next largest bundle set in the set member list. This process keeps repeating until a set is formed or there are no more sets in the set member list to test. If the members of the set member list have been tested, and no set has been formed, then the current bundle position is recorded as not belonging to a set. The loading map editor then determines if the user specified maximum for a set in the set member list has been violated. If so the editor performs a set recombination and ranking process. For example, if the number of sets of 2 exceeds the user specified maximum the editor does the following: For each set of 2, the loading map editor determines if another set of 2 forms a set of 4 meeting the symmetry requirements entered by the user. The loading map editor then determines the average attribute value and standard deviation for each newly formed potential set of 4 and ranks the potential sets of 4 based on minimum standard deviation. Next, the highest ranked sets (i.e., those with the lowest standard deviation) are assigned to the sets of 4 until the sets of 2 list does not exceed the maximum number allowed based on the user input. Those potential sets of 4 not assigned remain as sets of two. Next, the same process is performed to combine sets of 4 into sets of 8 assuming the user input parameters permit sets of 8 and the user specified maximum for sets of 4 has been violated. As a final step, the reinserted fuel bundles are assigned a priority number that, as described above, appears in the template. The fuel bundles positions are ranked based on (1) the attribute value for the fuel bundle in the associated position in the loading map if the fuel bundle position does not form part of a set; or (2) by the average attribute value of the set if fuel bundle position does form part of a set. A priority number is then assigned by this ranking with the fuel bundles having the same average attribute assigned the same priority number. This completes the template derivation process, the resulting template is then displayed in the loading map editor allowing the user to save the resulting template for future use. Using the present invention as described above, a core designer may capture his experience and rules of thumb associated with the initial design of a loading map. Furthermore, this knowledge may then be used by others to improve or adapt templates to existing core designs. Creating Loading Map The loading map editor according to the present invention includes additional functionality that allows the user to generate a loading map from the loading template. In addition, the loading map editor provides increased flexibility in creating the loading map by allowing the user the option of reloading fuel bundles currently residing in one or more fuel pools. After accessing, creating and/or editing a reactor core template using the loading map editor as discussed above, the user may then create a loading map using the template. From the file menu 30, the user chooses a “load” option. The loading map editor then displays a loading screen that includes a template access window, template information window, reload window and a load fresh window. The template access window provides a user with a drop down menu for selecting a loading template stored in the memory 18. The template information window displays summary information for the selected loading template. The summary information includes, but is not limited to, the number of fresh bundle types, the number of reinserted fuel bundle positions and the number of locked bundle positions in the loading template. The summary information may also indicate the number of fresh bundle types and number of reinserted bundles currently added in creating the loading map. FIG. 3 illustrates an exemplary embodiment of a reload window displayed by the loading map editor. The window is divided into two parts: a filtered fuel pool table 100 and a reloading pool 200. The filtered fuel pool table 100 lists (1) the exposed fuel bundles currently in the nuclear reactor under consideration, except for those fuel bundles in locked fuel bundle positions 38, and (2) the fuel bundles in one or more fuel pools for this and other nuclear reactors. As is well-known, exposed fuel bundles removed from a nuclear reactor are stored in what is known as a fuel pool. Fuel bundles from two or more nuclear reactor cores located at a same site may be stored in the same fuel pool. As shown in FIG. 3, the filtered fuel pool table 100 lists each exposed fuel bundle by its serial number and bundle name. Each fuel bundle is assigned a unique serial number, used to assure traceability of the bundle from a quality assurance perspective. The bundle name is a character string identifier used to identify the fuel bundle product line as well as nuclear characteristics, such as uranium and gadolinia loading. The filtered fuel pool table 100 also lists one or more attributes of each exposed fuel bundle listed. These attributes may include K infinity, exposure, and the last fuel cycle number for which the bundle was resident in the core. Additional attributes for an exposed fuel bundle may include: 1) bundle product line, 2) initial uranium loading, 3) initial gadolinium loading, 4) number of axial zones, 5) historical fuel cycle numbers previous to the most recent for which the bundle was resident in the core, 6) the corresponding reactor in which the fuel bundle was resident for each of the historical fuel cycles, 7) accumulated residence time, and 8) fuel bundle pedigree, a parameter that reflects the usability of the bundle for continued reactor operation. The fuel bundle pedigree is determined from a number of factors the foremost being an inspection of the fuel, either visually or by some other non-destructive test procedure, which is designed to detect a current failed fuel bundle or the vulnerability of the bundle to future failure. Failure mechanisms include such items as corrosion, debris impact, and mechanical bowing of the fuel bundle. Another factor affecting pedigree is possible reconstitution of a fuel bundle, which is a repair process involving the replacement of damaged fuel rods with replacement rods that may be a uranium containing fuel rod or alternatively, a non-uranium containing rod (e.g. stainless steel), henceforth referred to as a ‘phantom’ rod. A pedigree attribute might be ‘RU’ and ‘RP’ for reconstituted with uranium and phantom rods, respectively, and ‘DC’, ‘DD’ and ‘DB’ for damaged by corrosion, debris, and bow, respectively. A ‘blank’ pedigree attribute would designate a bundle that was undamaged and useable. All attributes with the exception of bundle pedigree are populated within the database via a direct relationship with the historical fuel cycles. The fuel pedigree attribute for non ‘blank’ designations are entered into the database via a separate process that is tied to fuel inspection and reconstitution services. In this process, the fuel bundles in a fuel pool are inspected and the pedigrees of the fuel bundles ascertained from the inspection. Then, a bundle status program is accessed. The bundle status program provides a GUI menu for ‘Fuel Inspection’, which is accessed by the user. The user clicks on the pulldown menu ‘Add’ from the ‘Fuel Inspection’ menu, and is presented with a pop-up for typing in the bundle serial number and the pedigree designation, such as ‘DD’ corresponding to a debris damaged bundle. The pedigree data entered in this manner is associated with the fuel pool database. The user may also click a ‘Census’ option from the ‘Fuel Inspection’ menu. Selecting this option will perform a query of the fuel pool database and present the user with a list of bundle serial numbers and corresponding attribute data, as described previously, for those bundles containing a non-null pedigree designation. The user may elect to change existing pedigree information by selecting the bundle entry, right-clicking a ‘Modify’ option, which activates the pedigree attribute field, and entering the modified pedigree information. For example, a bundle that was previously damaged may have been reconstituted. Alternatively, the user may right-click a ‘Delete’ option, which has the effect of reverting the bundle pedigree status back to null. The reloading fuel pool table 200 provides the same information for each fuel bundle as provided by the filtered fuel pool table 100. Additionally, the reloading fuel pool table 200 indicates the priority number 202 for each fuel bundle set as set forth in the loading template. As discussed above with respect to the loading template, reinserted fuel bundles may be assigned as a set of 1, 2, 4 or 8 bundles. Accordingly, FIG. 3 shows that the highest priority reinserted fuel bundle position(s) are a set of four fuel bundles, and the next highest priority reinserted fuel bundle(s) are a set of eight fuel bundles. The reloading fuel pool table 200 is populated by moving fuel bundles from the filtered fuel pool table 100 into the reloading fuel pool table 200. As further shown in FIG. 3, the reload window further includes a set of tools 120 for aiding the user in selecting and moving fuel bundles from the filtered fuel pool table 100 to the reload fuel pool table 200. The set of tools 120 include, but are not limited to, a filter tool 130, a move right tool 160, a move left tool 170 and a delete tool 180. A user selects the filter tool 130 by, for example, clicking on the filter tool 130. This opens a filter window as shown in FIG. 4. As shown, the filter window lists the same attributes listed in the filtered fuel pool table 100, and allows the user to indicate to filter based on the attribute by clicking in the selection box 132 associated with the attribute. When an attribute has been selected, a check is displayed in the associated selection box 132. The user may also unselect an attribute by again clicking in the associated selection box. In this case, the check mark will be removed. For each attribute, the filter window may display one or more filter characteristics associated with the attribute. For example, for the filter characteristics of the K infinity attribute, the user may select a filter operator 134 of greater than, less than, or equal to and enter in a filter amount 136 associated with the filter operator 134. As shown in FIG. 4, a user has selected to filter based on K infinity, chosen the greater than filter operator, and entered the filter amount of 1.2. As a result, the loading map editor will filter the fuel bundles in the filtered fuel pool table 100 to display only those fuel bundles having a K infinity greater than 1.2. As another example, the exposure attribute also has an associated filter operator and filter amount. As will be appreciated, the filter characteristics of an attribute will depend on the attribute. Also, as will be appreciated, other methodologies for indicating the filter characteristics may be possible. For example, for the cycle attribute, the filter window provides a drop down menu for selecting the cycle number. FIG. 4 shows cycles 2 and 4 selected from the drop down menu for the cycle attribute. As a result, the loading map editor filters the filtered fuel pool table 100 to display only those fuel bundles whose most recent residence was in cycle 2 or cycle 4. Similarly, the user may elect to filter bundles based on their pedigree, product line, etc. Once the attributes for filtering on have been selected and the filter characteristics have been entered, the user causes the loading map editor to filter the filtered fuel pool table based on this information by clicking on the OK selection box. Alternatively, the user may cancel the filter operation by clicking on the CANCEL selection box. The filtered fuel pool table 100 also provides a filtering mechanism for filtering the fuel bundles listed therein. A user may sort the filtered fuel pool table 100 in ascending or descending order of an attribute by clicking on the attribute heading in the filtered fuel pool table 100. Once the user clicks on the attribute, the loading map editor displays a popup menu with the options “Sort Ascending” and “Sort Descending”. The filtered fuel pool table 100 is then filtered in ascending or descending order of the attribute based on the option clicked on by the user. To move fuel bundles from the filtered fuel pool table 100 to the reload fuel pool table 200, the user selects the fuel bundles for transfer by clicking and dragging to highlight one or more of the fuel bundles in the filtered fuel pool table 100. Then the user clicks on the move right tool 160. This causes the selected fuel bundles to populate the highest priority unpopulated fuel bundle positions in the reload fuel pool table 200. Alternatively, a user clicks and drags the highlighted fuel bundles into one of the priority sections of the reloading fuel pool table 200. Fuel bundles may also be moved from the reload fuel pool table 200 back into the filtered fuel pool table 100 by selecting fuel bundles in the reload fuel pool table 200 and clicking on the move left tool 170. Alternatively, the selected fuel bundles may be clicked and dragged back to the filtered fuel pool table 100. The delete tool 180 provides the user with the function of deleting fuel bundles from either the filtered or reload fuel pool tables 100 and 200. The user may select one or more fuel bundles in one of the tables, and click the delete tool to delete the selected fuel bundles from the table. Next, the loading of fresh bundles into the template will be described. FIG. 5 illustrates an exemplary embodiment of a load fresh window displayed by the loading map editor. The window is divided into two parts: a fresh bundle types table 300 and a fresh bundle pool table 400. The fresh bundle types table 300 lists the available fresh fuel bundle types. As shown in FIG. 5, the fresh bundle types table 300 lists each fresh fuel bundle type by its bundle name. The bundle name is a character string identifier used to identify the fuel bundle product line as well as nuclear characteristics, such as uranium and gadolinia loading. The fresh fuel bundle types table 300 also lists one or more attributes of each fresh fuel bundle type listed. These attributes may include K infinity, fuel bundle product line, average uranium-235 enrichment, percent (as a function of total fuel weight) of gadolinia burnable poison contained in the fuel bundle, number of gadolinia-containing fuel rods, and number of axial zones, where an axial zone is defined by a cross-sectional slice of the bundle that is homogeneous along the axial direction. Other attributes of the fresh bundle may include parameters for predicted thermal behavior, such as R-factors and local peaking, calculated for various bundle exposure values. R-factors are used as inputs to the critical power ratio (CPR) and are determined from a weighted axial integration of fuel rod powers. Local peaking is a measure of the fuel rod peak pellet and clad temperature. The fresh bundle pool table 400 provides the same information for each fuel bundle as provided by the fresh bundle types table 300. Additionally, the fresh bundle pool table 400 indicates the type number 402 for each type of fresh bundle in the loading template and then number of fresh fuel bundles of that type in the loading template. FIG. 5 shows that the first type of fresh fuel bundle position(s) are a set of four fuel bundles, and the next type of fresh fuel bundle(s) are a set of eight fuel bundles. The fresh bundle pool table 400 is populated by moving fuel bundles from the fresh bundle types table 300 into the fresh bundle pool table 400. As further shown in FIG. 5, the load fresh window includes the same filter tool 130, move right tool 160 and delete tool 180 for aiding the user in selecting and moving fuel bundles from the fresh bundle types table 300 to the fresh bundle pool table 400 as already described above. As will be appreciated, because the attributes for the fresh fuel bundles are different than the reinserted fuel bundles the filtering characteristics may also differ accordingly. The loading map editor also provides, as shown in FIG. 5, for filtering the fresh bundle types table 300 in ascending or descending order of an attribute in the same manner that the filtered fuel pool table 100 may be sorted. The selection and moving process for fresh fuel bundles does differ from the process for moving burnt fuel because the destination of the fuel must be chosen in the set fresh fuel bundle pool table 400 located on the right side of the fresh bundle types table 300. Namely, after a user selects the fresh bundle type from the fresh bundle types table 300, the user then selects one or more fuel bundle positions in the fresh fuel bundle pool table 400. By selecting the move right tool 160, the selected fuel bundle positions in the fresh fuel bundle pool table 400 are populated with the selected fresh bundle type. Alternatively, the user may click and drag the bundle type into the fresh fuel bundle pool table 400. Unlike with the filtered fuel pool table 100, the fresh fuel types are not removed from the fresh bundle types table 300 but are, instead, copied as fuel bundles into the fresh bundle pool table 400. Once the reinserted and fresh fuel bundle positions 38 are filled using the tools described in detail above, the user may click on a “Load” button 184 displayed in the loading screen. This step performs the assignment of the fresh and reinserted fuel bundles from the fresh bundle pool list 400 and reloading pool 200, respectively, to the physical I,J locations in the core loading map. The core loading map is then displayed to the user as shown in FIG. 7. The user may then save the created loading map by choosing “Save” or “Save As” from the Options Menu 760 within FIG. 7. Further refinement of the core loading map may be performed. The user has the option of performing a “swap” or exchange of two bundles, which may be between either fresh or reinserted bundles. The user enters “Shuffle” by clicking on a “shuffle” radio button 707 under “Mode”. The user then selects a symmetry 710 followed by selection of a first and second location within the core loading map. All symmetric partners of the first location will exchange with all symmetric partners of the second location. While in shuffle mode, this process may be repeated to perform any number of rearrangements of fresh and reinserted fuel within the core loading map. A second refinement to the core loading map involving a change in the fresh fuel bundle design (i.e. a change in fresh fuel type or IAT, a number assigned to a specific fuel bundle design within the loading map) may be performed. The user enters “IAT” mode by clicking on a “IAT” radio button 709. The IAT is a designation of a fuel bundle type (e.g., IAT 1, IAT 2, etc . . . ). For example, a given IAT number may be reserved for fresh fuel bundles. The user selects a symmetry and then clicks on a fresh location in the core loading map. A pop-up window appears with an IAT pull-down, from which the user selects a new IAT for the targeted location (along with the symmetric partners of the targeted location). While in IAT mode, the process of changing fresh bundle types within fresh core locations may be repeated any number of times. Having created the loading map, the user may then perform simulations on reactor core performance, etc. using the loading map created according to the methodologies of the present invention. The “Shuffle” and “IAT” modes for refining the loading map may be performed iteratively with the simulating step, with the results from the simulation being used as a guide to facilitate improvement to the core loading map. By allowing the user to draw on the resources of the fuel pool(s), the present invention provides for greater flexibility in the creation of the loading map and may also reduce the overall cost in loading a nuclear reactor core. Creating Bundle Groups Once a loading map has been generated (e.g., with the above described methodology), a core designer may assign fuel bundles into one of a plurality of bundle groups. The bundle groups are generated with attributes that define the bundle group. FIG. 6 illustrates a flow chart of a process for creating bundle groups. Referring to FIG. 6, in step S600, a core designer enters a group mode within an input deck of a loading map editor. This step will be described with reference to FIG. 7. FIG. 7 illustrates a screen shot of a portion of the fuel bundles within a loading map according to the methodologies of the present invention. As shown, the loading map includes a plurality of fuel bundles 720. The position of the fuel bundles 720 as viewed in the loading map relates to the position of the fuel bundles 720 in the core. Further, each of the fuel bundles 720 includes displayed data associated with the fuel bundles 720. For example, as shown in FIG. 7, the displayed data includes a bundle group number (in the top left portion of fuel bundles 720 associated with a given bundle group), bundled average K infinity (K-inf) (not shown) (e.g., in the center portion of fuel bundles 720), and the exposure in GigaWatt Days/short ton (GWD/st) (not shown) (e.g., in the bottom portion of fuel bundles 720). As shown, a core designer utilizing the above-described loading map editor enters a Group Mode by selecting a tab “Bundle Groups” 700 from a plurality of tabs 703. Then, the core designer selects group mode 705 from a plurality of mode options (e.g., group 705, shuffle 707, IAT 709). Referring to FIG. 6, in step S605, the core designer determines whether to execute a grouping operation manually or with a Wizard. The Wizard is a software tool which may be used to automate bundle group generation. If the core designer decides to execute the grouping operation manually, the core designer may assign fuel bundles to bundle groups through a bundle by bundle, iterative process and the process advances to step S610. In this embodiment, the core designer analyzes each fuel bundle individually and determines to which bundle group the fuel bundle under consideration is to be assigned based on at least one of a plurality of attributes. The criteria the core designer may consider when generating or editing bundle groups will be described later. In step S610, the core designer chooses between performing a grouping operation on an existing bundle group (or individual fuel bundle belonging to a bundle group) or perform a grouping operation on an existing bundle group, then in step S613, the core designer selects a bundle group on which to execute a grouping operation. The core designer left clicks a fuel bundle belonging to the selected bundle group. Referring again to FIG. 7, fuel bundles 720 represent a sample of the available fuel bundles which the core designer may click to designate a bundle group. Fuel bundles 722 represent fuel bundles belonging to a bundle group selected by the core designer. As shown, the selected fuel bundles 722 are highlighted and each belong to the same group (i.e., group “1”). The group number for group “1” is displayed by group designation number 724. Group designation number assignment will be described later. Thus, group “1” is set to an active state, which means that grouping operations will affect group “1” until the core designer sets another group to the active state. Then, in step S615, the core designer right clicks on a selected fuel bundle and a list of available grouping operations associated with the fuel bundle become available. If the selected fuel bundle belongs to a group, the available grouping operations include 1) Remove Bundles From Group and 2) Delete Group. If the selected fuel bundle does not belong to a group, the available grouping operations only 3) Add Bundles to Group. The grouping operation of 4) Create Group is not available here and will be described later with respect to steps proceeding from step S635. The core designer selects one of the available grouping operations and left clicks the selected operation with the mouse. If 1) Remove Bundles From Group is clicked, the process advances to step S620. If 2) Add Bundles To Group is clicked, the process advances to step S625. If 3) Delete Group is clicked, the process advances to step S630. In step S620, the selected fuel bundle and its symmetric partners (e.g., the fuel bundles associated with the selected fuel bundles based on the symmetry being 1, 2, 4M, 4R, 8, etc.) are removed from their bundle group and reset to a status of not belonging to a group. Thus, the group designation number in the upper left portion of the fuel bundle is removed. The process may then advance to step S655. In step S625, the selected fuel bundle and its symmetric partners are added to the group selected in step S613. The group designation number in the upper left portion of the selected fuel bundle and its symmetric partners is set to match the group designation number of the selected group. The process may then advance to step S655. In step S630, the group selected in step S613 is deleted entirely. Thus, each fuel bundle in the loading map belonging to the selected group is reset to a status of not belonging to a group. The group designation number for each fuel bundle in the group is removed from the upper left portion of the fuel bundle. The process may then advance to step S655. In step S655, the execution of the grouping operation is completed. If another grouping operation is required, the process advances back to step S605. If another grouping operation is not required, the process advances to step S680. In step S680, no additional steps are required and all operations required by the core designer are complete. The core designer may select another of the plurality of tabs to enter another feature of the loading map editor or exit from the loading map editor. Returning to step S610, if the core designer chooses to create a new bundle group, then in S635, the core designer selects a desired bundle group symmetry (e.g., 1, 2, 4M (Mirror), 4R (Rotational), and 8). The core designer left clicks the selected bundle group symmetry. Referring to FIG. 7, a symmetry 710 of 8 is selected by the core designer. The process may then advance to step S640. In step S640, the core designer selects a first fuel bundle to form the new bundle group. In one embodiment, the core designer may left click on the New tab. The core designer then left clicks on a fuel bundle not belonging to a bundle group in step S650. The fuel bundle is added to the new group and given a group designation number. The group designation number is the lowest unique natural number available for group designation. For example, if groups “1”, “2” and “3” currently exist in the loading map, the new group would be designated as group “4”. In this example, the number “4” would appear in the upper left portion of each fuel bundle assigned to group “4”. In another example, if groups “1”, “3” and “4” currently exist in the loading map, the new group would be designated as group “2”. In this example, the number “2” would appear in the upper left portion of the fuel bundle assigned to group “2”. Alternatively, in another embodiment of step S650, instead of left clicking the New tab, the core designer right clicks a fuel bundle not belonging to a bundle group. Available grouping operations become available to the core designer. The only available grouping operation for the fuel bundle not belonging to a bundle group is 4) Create Group. The core designer left clicks on the 4) Create Group operation. The fuel bundle is then added to the new group as described above with respect to the New tab embodiment. After the first fuel bundle is assigned to the new group, the process advances to step S645. In step S645, the core designer decides whether more fuel bundles will be added to the newly created bundle group. Generally, bundle groups are designated as either exposed or fresh. Fresh bundle groups include only fresh fuel bundles. Exposed bundle groups may include fresh fuel bundles. However, in “mixed” bundle groups, each of the included fuel bundles are treated as exposed (e.g., with respect to swapping operations) regardless of whether the includes fuel bundles are fresh or exposed. In one embodiment, at least three fuel bundles per symmetry section are required in an exposed bundle group since this is the minimum number of fuel bundles required for a swapping operation only including fuel bundles within the symmetry section, which will be described later. If the core designer chooses to add an additional fuel bundle to the newly created bundle group, the process returns to step S650. Otherwise, the process advances to step S655. Returning to step S605, if the core designer chooses to use the Wizard, then the core designer left clicks on the wizard button shown in FIG. 7 in step S660. FIG. 8 illustrates a screen shot of a Wizard window according to the methodologies of the present invention. Referring to FIG. 8 with respect to step S660 of FIG. 6, when the core designer left clicks on the wizard button 740, a window 905 including an Add to Group column 910, an IAT column 915, a Fresh Column 920, symmetry options 925, and a Create Group button 930 become available to the core designer. The 1) Add to Group column 910 includes input tabs and the 2) IAT and 3) Fresh columns 915/920 include indicators to the core designer. The 2) IAT column 915 displays all available IATs to the core designer. The 3) Fresh column 920 indicates whether the IAT in an associated row is fresh. If the indicator is checked (e.g., indicator 932), the IAT is fresh. If the indicator is not checked (e.g., indicator 934), the IAT is not fresh. Returning to FIG. 6, in step S665, the core designer selects a desired bundle group symmetry (e.g., 1, 2, 4M (Mirror), 4R (Rotational), and 8) from the available symmetry options. The core designer left clicks the desired bundle group symmetry 925 and the process advances to step S670. In step S670, the core designer selects IAT types (i.e., bundle types) to add to the new bundle group. The core designer left clicks the input tabs in the Add to Group Column 910 in the same row as the desired IAT type. For example, the core designer may left click on the input tab 936 associated with a fresh IAT type (e.g., the fresh indicator for that row is checked). Thereafter, only additional input tabs associated with fresh IAT types may be added to the new bundle group with the Wizard. However, either fresh or exposed fuel bundles may still added manually. Alternatively, if the core designer initially selects an input tab 938 that is not associated with a fresh IAT type (e.g., the fresh indicator for that row is not checked), only additional input tabs not associated with fresh IAT types may be added to the new bundle group with the Wizard. However, either fresh or exposed fuel bundles may still be added manually. After the core designer completes selection of IAT types to be added to the new group, the process advances to step S675. In step S675, the core designer clicks on the Create Group button 930 within the window 905 of the Wizard tool. Each fuel bundle associated with the selected IAT types is added to the new group and given a group designation number. If a fuel bundle associated with one of the selected IAT types already belongs to another bundle group, the fuel bundle is removed from the other bundle group and is added exclusively to the new group. In one embodiment, if less than all of the fuel bundles of a given IAT type are desired in the new bundle group, the core designer may use the Wizard to add all fuel bundles of the given IAT type to the new group. Then, the core designer may remove the fuel bundle(s) not desired in the new group manually using above-described methods. As discussed above, the group designation number is the lowest available unique natural number available for group designation. For example, if groups “1”, “2” and “3” currently exist in the loading map, the new group would be designated as group “4”. In this example, the number “4” would appear in the upper left portion of each fuel bundle assigned to group “4”. In another example, if groups “1”, “3” and “4” currently exist in the loading map, the new group would be designated as group “2”. In this example, the number “2” would appear in the upper left portion of each fuel bundle assigned to group “2”. The process then advances to above-described step S655. As discussed earlier, bundle groups are divided into exposed bundle groups and fresh bundle groups. Bundles within an exposed bundle group may “shuffle” or be exchanged with other bundles within the same bundle group. In an example optimization algorithm, a minimum of three sets of symmetric bundles may define an exposed bundle group, the exposed bundle group including exposed fuel and/or fresh fuel. In contrast to exposed bundle groups, fresh bundle groups include only fresh fuel. Bundles within the fresh bundle group are not exchanged or shuffled, as described above with respect to the exposed bundle group, but rather may be modified through a bundle characteristic (e.g., an IAT type) selection. The core designer selects a list of allowable IAT types for a given fresh bundle group from among a list of all available IAT types for the given fresh bundle group. While step S670 of FIG. 6 as described above shows how IAT types may be selected for inclusion in a new bundle group, an example will now be given where IAT types in an existing fresh bundle group in a current loading map may be modified. Referring to FIG. 7, the user enters IAT Mode 709 by left ‘clicking’ on the IAT radio button 709. The core designer then left ‘clicks’ on a fresh bundle within a previously defined fresh bundle group 722 to reveal a pop-up window as shown in FIG. 9, which displays to the core designer a list of available IAT types 950 for the selected fresh bundle group 722. Referring to FIG. 9, the core designer may select or de-select among available IAT types 950 with corresponding bundle names 955 and checkboxes 960/965. The core designer selects or de-selects an available IAT type 950 by left clicking on the checkbox 960/965, wherein checkboxes 960 are shown de-selected and checkbox 965 is shown selected. For a current loading map, certain checkboxes 960/965 may be selected automatically (e.g., not by the core designer), thereby indicating IAT types already included in the selected fresh bundle group 722. An optimization of the selected fresh bundle group 722 may be performed if at least two IAT types are selected. Modifying the IAT type of a fresh bundle may change the distribution of enriched uranium and gadolinia within the fresh bundle design. Criteria for Bundle Group Assignment Examples of criteria the core designer may consider in assigning fuel bundles to bundle groups will now be described. In one example, the core designer may desire all fuel bundles with certain IAT types to be included within a bundle group. The core designer may use above-described methods (e.g., manually creating the bundle groups or using the Wizard tool) to create the desired bundle groups. In another example, consider a situation where the core designer requires that no fresh fuel bundles be loaded into a bundle position adjacent to a control blade (i.e., referred to as a “controlled location”). In this case, the core designer creates a bundle group only including fuel bundles surrounding the controlled locations. The core designer further sets an attribute for the bundle group being that only once burnt fuel bundles (i.e., bundles that have been in the core for only one cycle of operation) may be loaded into the bundle group (e.g., by adding only once burnt fuel bundles with the above-described manual process). During an optimization process, the core designer will only consider proposed loading maps which meet the criteria established for each of the bundle groups. Thus, the core designer uses the attributes which define a bundle group as a constraint, which will be discussed in further detail below. Bundle Groups as a Constraint in an Optimization Process An example implementation of the assigned fuel bundle groups as a constraint in an optimization process will now be described. As described above, fuel bundles within a core may be assigned to one of a plurality of bundle groups. Collectively, the fuel bundle assignments constitute a loading map. During the cycle of operation in a nuclear reactor, the loading map does not change. However, a period between cycles of operation (referred to as an “outage”), reserved for general maintenance of the reactor, may include a repositioning of fuel bundles within the core. The fuel bundles repositioned during the outage may be designated in a selected loading map for a next cycle of operation. The selected loading map may be one of a plurality of loading maps evaluated. The selection may be based on a trial and error process by a core designer and/or by an evaluation result of an objective function. The objective function methodology is disclosed in U.S. patent application Ser. No. [10/246,716], titled METHOD AND APPARATUS FOR EVALUATING A PROPOSED SOLUTION TO A CONSTRAINT PROBLEM, by a subset of inventors of the subject application. The established bundle groups may serve as a constraint by limiting the number of proposed loading maps to be considered (e.g., by either the core designer and/or the objective function). The constraint is that fuel bundles may only be exchanged with other fuel bundles of the same bundle group within the same region of symmetry. The region of symmetry may refer to any of the above-described symmetry regions, including a quadrant (e.g., a quarter of the entire core as implemented in the United States), a semi (e.g., half of the entire core as implemented in Europe), an octant, etc. Since fuel bundle exchanges are often the “bottleneck” of the outage (e.g., the duration of the outage may scale with the number of fuel bundle movements), the duration of the outage may be reduced by limiting exchanges between fuel bundles (e.g., within the region of symmetry) to fuel bundles within the region of symmetry since these exchanges require less time. Thus, before evaluating a proposed loading map, proposed loading maps including violations of the above-described bundle group constraints are culled and removed from consideration. Then, the reduced set of proposed loading maps may be configured for inclusion within an objective function for evaluation by the core designer. Therefore, evaluation time (e.g., by the core designer or the objective function) required to select the loading map for the next cycle of operation may likewise be reduced since fewer loading maps are considered. The invention being thus described, it will be obvious that the same may be varied in many ways. Such variations are not to be regarded as a departure from the spirit and scope of the invention, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the invention.
047012803	abstract	The invention provides for a procedure for permanently storing radioactive material in a rock chamber, comprising an enclosed construction (4) completely separated from the walls, floor and ceiling of the rock, made of permanent material impervious to water, in which materials in encapsulated (9) form are stored, and the outer spacing between the walls, ceiling and floor of the rock and the construction is completely filled with material not impervious to water (5). In the invention, the encapsulated material is placed inside the construction by means of a robot (10) on both sides of a track provided for the robot. When the construction is totally or partially filled with material, the construction or part of it containing material is sealed with bentonite or similar, leaving a passage (13) for the robot in the construction, which passage is filled with water-displacing material that is easy to dig. When the construction is completely filled, it is sealed up and the robot (10) is lifted up through a lift-shaft (3) connecting the construction with ground-level. The lift-shaft is sealed at least at the level of the construction and at ground-level.
	abstract	The invention relates to a method and apparatus of control and/or determination of energy and/or velocity of a beam of a charged particles extracted from a synchrotron, the synchrotron using a radio frequency field to redirect the charged particles through an extraction material. The method and apparatus for extracting a group of charged particles at a known and controlled energy is optionally used in conjunction with any apparatus and/or technique coupled to a synchrotron, such as a charged particle cancer therapy system. For example, the knowledge/control of the charged particle beam delivery energy is used in conjunction with a multi-axes and/or a multi-field tumor irradiation system in combination with a tumor treatment plan to accurately ablate or irradiate a tumor.
050858254	summary	FIELD OF THE INVENTION This invention relates to water cooled nuclear fission reactors of the so-called boiling water type. Boiling water nuclear reactors comprise a steam generating plant wherein reactor water coolant is circulated through a core of heat producing fissionable nuclear fuel to transfer thermal energy from the fuel to the coolant and produce steam. The steam is then used to drive turbines and other machinery employing steam, such as for electrical power generation. BACKGROUND OF THE INVENTION Due to the prodigious quantities of thermal energy produced by fissioning nuclear fuel, it is imperative to maintain the fuel core of commercial water cooled nuclear fission reactors submerged within heat transferring coolant water. The conveyance of heat out away from the energy producing fuel core by means of a fluid coolant is needed to preclude the possibility of hazardous conditions or reactor damage such as could occur with an overheating meltdown within the fuel core unit. Such a potentially destructive occurrence can result from a loss-of-coolant accident (LOCA) caused by an extensive breach of a major reactor coolant receptacle or conduit. To cope with this hypothetical accidental event, commercial water cooled nuclear fission reactors are provided with large reservoirs of water and dedicated safety injection systems(s) capable of supplying supplementary coolant water to the reactor vessel for cooling the fuel core and maintaining lower or normal operating temperatures. Typically, automatic safety measures are employed to activate and operate systems for supplying this supplementary coolant water as needed to the fuel core to replace or supplement the original coolant water lost due to some mishap. A different but additional safety measure commonly employed in commercial water cooled nuclear fission reactors comprises means to deal with any failure of the reactor control rods to effectively perform their designed fission regulating function. Nuclear reactor control rods, comprising elongated units containing neutron absorbing material such as boron or a compound thereof, are designed to be reciprocally movable into and out from the fuel core of fissionable material. The level of fission activity of the fuel material in a core of a reactor, and in turn heat produced, is determined or controlled by the amount of neutron absorbent advanced into or withdrawn from the fuel core with the control rod units. Moreover, the fission reaction of the plant can be rendered subcritical or terminated by inserting sufficient neutron absorbent material housed within control rods into the fuel core to deprive the fuel of the needed quantity of fission produced neutrons for maintaining the self perpetuating fission reaction, or so-called chain reaction, and in turn heat produced. However, in the event that the control rods fail to perform their intended role of governing and/or ceasing the fission reaction for any cause, mechanical, electrical or personnel malfunction, an auxiliary backup system is frequently provided for depriving the fuel core of the neutrons essential for maintenance of the heat producing fission reaction. Commonly this system comprises a supply of a water solution of a soluble boron or gadolinium compound, or so-called poison, which when introduced into the nuclear reactor vessel, merges with the reactor coolant and thus pervades the fuel core region to absorb the fission produced neutrons needed for continuing the fission reaction. Thus, in the remote event of any control rod failure, the poison solution is fed from a reservoir into the reactor vessel by suitable means on a signal of problem whereupon the fission reaction is terminated. A typical arrangement in commercial water cooled nuclear fission reactor plants for incorporating the foregoing standby safety systems which inject auxiliary coolant water to temper the fuel core temperature, or inject a neutron absorbing solution to deprive the fuel of neutrons, utilized an apt gas, such as nitrogen, for a propellant to drive the liquid water or boron solution from its source or reservoir through communicating conduits into the reactor vessel. Thus, auxiliary coolant water or poison solution is maintained within a closed vessel or tank under sufficient gas pressure to drive the liquid contents into the reactor vessel through an appropriate arrangement of conduits upon a manually or automatically actuated signal responding to a malfunction. SUMMARY OF THE INVENTION This invention comprises an improvement in a standby safety injection system for water cooled, nuclear fission reactor plants. The improvement of the invention enhances the injection delivery means, and includes a connecting of the reservoirs of two safety systems to provide an alternative auxiliary source of propelling gas for each system. Additionally this invention comprises a unique arrangement of control valves and fluid conduits joined with sensing means for initiating the standby safety injection units of either system into appropriate action for dealing with a reactor emergency or malfunction. OBJECTS OF THE INVENTION It is a primary object of this invention to provide an improved dual liquid standby safety injection system for nuclear fission reactor plants. It is an additional object of this invention to provide a liquid standby safety injection system for water cooled nuclear fission reactor plants having an alternative auxiliary source of propelling gas. It is a further object of this invention to provide a multiple liquid standby safety injection system for water cooled nuclear fission reactor plants having an improved gas propellant and application system that enhances performance as well as economic benefits. It is a still further object of this invention to provide a liquid standby safety injection system for water cooled nuclear fission reactor plants having an improved arrangement of fluid conveying conduits and flow controlling valves. It is also an object of this invention to provide improved multiple liquid standby safety injection systems for water cooled nuclear fission reactor plants providing for alternate application of coolant or neutron absorbent with a common source of propellant gas selectively applied or transferred, coupled with a unique arrangement of fluid transferring conduits and flow controlling valves combined with operating sensors.
050808583	claims	1. A fuel assembly in the form of an elongated multi-corner channel having a wall, said channel being connected to a coolant intended to traverse the channel, a bundle of similarly elongated fuel rods arranged in the channel and retained by a plurality of spacers placed along the bundle and comprising a number of cells surrounded by an outer frame which is formed of a band standing on edge, wherein said band on an upstream side of the spacer is extended by a skirt in which openings are arranged, said skirt making close contact with said wall, deflection fins being arranged in certain openings to deflect a coolant flowing along said wall in a direction towards the center of the respective spacer. 2. A fuel assembly according to claim 1, wherein the skirt is arranged open in the above-mentioned corner and makes contact in a resilient manner with said wall. 3. A fuel assembly according to claim 1, wherein the spacer frame is provided with windows in which deflection fins are arranged in order to strengthen the deflecting effect of the fins provided in the skirt. 4. A fuel assembly according to claim 1, wherein each side piece of the skirt, corresponding to one side of the multi-corner channel, has an edge facing the direction of flow, which edge in a direction against the flow has been give a substantially arrow- or convex-like shape. 5. A fuel assembly according to claim 1, wherein each side piece of the skirt is made in a separate piece and provided on the downstream side with members to be connected with the frame of the spacer. 6. A fuel assembly according to claim 5, wherein the skirt is made of Zircaloy. 7. A fuel assembly according to claim 5, wherein the side pieces of the skirt are connected to each other at the corners so that the skirt forms a sleeve. 8. A fuel assembly which comprises a wall that defines an elongated, multi-cornered channel through which coolant can flow from an upstream end to a downstream end, a bundle of elongated fuel rods positioned in said channel, and a plurality of spacer means positioned at separated locations along said bundle of fuel rods, each of said spacer means comprising an elongated outer band, means defining a plurality of cells within the outer band and a center line, and a skirt which extends from said outer band toward said upstream end of said channel and in contact with said wall, said skirt having a plurality of openings therein and including fins located upstream of some of said openings, said fins extending inwardly of said wall and in said downstream direction so as to deflect coolant flowing in said downstream direction inwardly of said wall and towards said center line.
055457983	description	DESCRIPTION OF THE PREFERRED EMBODIMENT Preliminary Comments: These figures show the preferred treatments of radioactive ion-exchange resins. The embodiments described are consistent with treatments now used for ion-exchange resins contaminated by water carried by alloyed metal tubing through certain nuclear reactors at power stations. For teaching the method of this invention, the figures show chemically or physically important stages of the treatments: For each figure, on a local scale the stage order indicated is substantially followed, although the stage times may be almost simultaneous, as is discussed later. On the bulk scale the stages are reached at different times as the solid resin depolymerizes over a period of time because only the resin surface is exposed to reaction. FIG. 1: This figure shows the preferred embodiment for treatment to reduce the burial volume for radioactive ion-exchange resin when barium hydroxide is the anchor material supplied. FIG. 1 Stage 1: A sealable container with stirrer is supplied an aqueous slurry of barium hydroxide anchor material and radioactive ion-exchange resin, e.g., from operations of a BWR nuclear-electric power station. Barium anchoring ions, Ba.sup.++, load hydrogen ion sites of the ion-exchange resin, thereby anchoring the cation-exchange groups, here sulfonate groups, and decaying atoms, which may be on the resin or in aqueous solution. First-treated resin is formed. Excess anchoring ions also remain, and water is drained off for recycle to more aqueous slurry. FIG. 1, Stage 2: By heating the stirred container to a suitable temperature, e g., to 150.degree. C., water vapor is driven off and is collected for recycle to the BWR turbine generator. FIG. 1, Stage 3: By further heating toward 300.degree. C., a series of reactions take place. (i) Heat and further anchoring ions, perhaps with the assistance of water, attack the bonds between the anchored cation-exchange groups, here sulfonate groups, and the organic portion of the first-treated resin; the attack converts sulfonyl groups to firmly bonded radioactive material such as radioactive BaSO.sub.4, i.e., synthetic barite mineral, that is at least in part chemically freed from organic material. (ii) The attack also releases organic polymer residue that is at least in part freed from anchored sulfonate groups and their attached decaying atoms. (iii) The heated organic polymer residue is also allowed to depolymerize, at least in part, and barium compounds may catalyze the depolymerization. (iv) Vaporization of the depolymerized resin allows the organic material to be removed from the firmly bonded radioactive material and be collected elsewhere. FIG. 1, Stage 4: The condensed organic vapor may need final purification, e.g., washing with dilute acid to remove contaminants such as traces of radioactive material or hazardous material or trimethyl amine from anion resin that may have been present. FIG. 1, Stage 5: The firmly bonded radioactive material goes to storage or burial, and the condensed organic material goes to nonradioactive disposal. FIG. 2, Stage 1: The system is supplied radioactive ion-exchange resin that has been roughly dried consistent with power station policy, e.g., by squeezing it and pumping vapor from it. This resin is placed in a separation contain-along with bonding material (also called anchor material) which, in this preferred embodiment, is a mixture of sodium hydroxide and potassium hydroxide. Aqueous hydroxides form immediately. Other materials comprising oxides could also be used in powder or liquid form, or in other form which could make firmly bonded radioactivity of the next stage. FIG. 2: This figure shows the earlier preferred embodiment to reduce the burial volume for radioactive ion-exchange resin when sodium hydroxide-potassium hydroxide is the bonding (i.e., anchor) material supplied. FIG. 2, Stage 2: The hydroxide solution brings strong ionic environments around both the exposed radioactive ions and the ion-exchange structures attached to the resins. Many surface radioactive ions will move into the hydroxide-solution region--there the radioactive ions are surrounded by ionic fields which bond them more firmly than nonionic organic regions of the resin can do it. Also, the inorganic ion-exchange groups bonded to the organic resin become subject to strong bonding from the ionic aqueous phase. If thermal agitation breaks organic bonds, the originally ion-exchange groups will remain with an ionic aqueous phase or other largely ionic phase. Ion exchange will lead to some removal of interior decaying atoms out to hydroxide solution. However, completing Stage 2 will require conversion of the resin to a different form which gives the hydroxide access to the interior of the solid resin. Heating and various decomposition stages as follow are used to give that access. FIG. 2, Stage 3: Heating of the hydroxide-resin mixture is carried out in a portion of the separation container. The heating, assisted by catalytic and chemical action of the hydroxide, causes (i) depolymerization of much of the resin to form organic liquid solution which is largely immiscible with water, (ii) separation of much of the decaying atoms and much of the ion-exchange portion the resin into aqueous ionic solution, and (iii) formation of some resin residue mixed with some trapped decaying atoms, which mixture is immiscible with either the aqueous or the organic phase. FIG. 2, Stage 4: In this embodiment the physical separation of the decaying atoms from the organic material is primarily by vapor transport. The vaporization and subsequent condensation in another region of the separation container moves major portions of the nonradioactive material where it can be collected and be moved on toward disposal. The vapor transport is assisted by water vaporization with condensation at a collection region of the separation container. The steam acts as a carrier gas (steam distillation). Other carrier gases can also be used for transport of organic vapor to the collection. If the separation container is hermetically sealed, reduced system pressures can assist the vapor transportation. The reduced pressures lower the boiling points for the vapors evolved, and vapor transport is sharply increased by boiling. While vaporization is preferred, in come cases other techniques such as aqueous-organic solvent extraction may also usefully be used. FIG. 2, Stage 4A: The vaporized organics are condensed and held for further vapor condensation as a result of other techniques. FIG. 2, Stage 4B: Here material not decomposed by depolymerization is subjected to pyrolysis by heating. Some pyrolysis is essentially inevitable as corollary to the heating for depolymerization. The depolymerization and pyrolysis in some ways blend into one another: However, the depolymerization refers more to breaking the bonds formed by the original polymerization of reactants, while pyrolysis refers more to breaking miscellaneous bonds, as in charring paper. FIG. 2, Stage 5: Here carbonaceous material, carrying the hydroxide residues, has now largely altered chemically. FIG. 2, Stage 6: Material from Stages 4 and 5 may be combined. They move separately or together to monitoring for possible environmental contaminants. FIG. 2, Stage 8: The nonradioactive organics are monitored. If they pass the monitoring they are ready for release, possibly to recycle and possibly to nonradioactive disposal. FIG. 2, Stage 9: The radioactive material goes to radioactive disposal in smaller volume than it would have had in current technology. FIG. 2, Stage 10: The nonradioactive material, in this case free of chemical hazards as well, is disposed of or is recycled. FIG. 3: This figure shows how the essentials of this preferred embodiment in FIG. 2 may be usefully be expanded or altered. All stages retain their meanings as in FIG. 2. Primarily the stages not included in FIG. 2 are discussed below: FIG. 3, Stage 4C: The carbonaceous material and decaying atoms which might have moved to disposal may also be oxidized primarily to carbon dioxide, but moisture and other molecules may be released during oxidation. This oxidation can remove most of the remaining carbon, but inorganics such as oxides, hydroxides, carbonates, sulfates, etc., will remain, holding the decaying atoms. FIG. 3, Stage 4D: Other techniques may be used instead of vaporization to separate radioactive and nonradioactive portions of the original radioactive ion-exchange resin. For example, as the material sits after depolymerization and corollary initial polymerization, three regions at least will be present, i.e., a liquid organic phase, a liquid aqueous phase, and solid residuals from the depolymerization. In effect, a rough solvent extraction already has been achieved by the depolymerization. The separation already may be adequate to provide easy separation of radioactive and nonradioactive materials. Radioactive aqueous liquid can be poured off and dried with radioactive solids then move in small volume to radioactive disposal. And organic liquid decanted before drying off the water can move to nonradioactive disposal. FIG. 3, Stage 7: Once the larger organic molecules are condensed, nonradioactive carbon dioxide can be collected at another collection site in a separation container. The two sites are not distinguished in the figure but they normally will be separate. Production of gases such as carbon dioxide should be minimized in early stages of the resin destruction to avoid producing large amounts of gases which are difficult to collect and monitor before they are prepared for disposal. By conceptual design, residual carbonaceous chars will be in relatively small amounts and may be oxidized to carbon dioxide. This and other gases may be collected and concentrated in several ways, e.g., (i) with cooling at lowered temperatures and at pressures higher than atmospheric, (ii) by low-temperature sorption, (iii) by collection on chemical scrubbers, or (iv) or by combinations of ways. Carbon dioxide is collected and held in a concentrated form. Therefore, simple analyses can be given enough time and sufficient concentration of decaying atoms to assure accurate measurements. The environmentally benign collected gas can be released to the atmosphere. Experimental Case 1 A typical case with the preferred embodiment using barium hydroxide anchor material proceeded as follows: First, solid UF.sub.4 was contacted for 15 minutes with fresh, sulfonated polystyrene cation-exchange resin in water, thereby adding a distinct U.sup.++++ color to the resin. Next, the wet resin was mixed with enough Ba(OH).sub.2 anchor material in slurry form to allow ultimate formation of BaSO.sub.4 from all the sulfonyl groups in the resin present. This mixture was stirred occasionally for a half hour, allowed to settle, and freed of much of the water by decantation. The wet mixture of anchor material, radioactive resin, and some solid UF.sub.4 was put into a sealed borosilicate-glass system with provision for displacing air, evacuating, heating, and vaporizing and condensing both water and volatile organic materials. The water was largely dried away, either by partial evacuation or by flow of carrier gas, with vapor collection in either case. Consideration of the experimental behavior and theoretical objectives leads the inventor to conclude that anchoring ions had attached to sulfonyl groups and anchored them. At this point one Ba.sup.++ attached to two sulfonyl groups, and hydrated barium hydroxide was also present. Later analyses showed the water to be substantially free of decaying atoms. The system was further heated toward 300.degree. C., again with partial evacuation or use of argon carrier gas to sweep organic vapors to condensation sites. Fog from vaporization of large organic molecules became increasingly evident as heating proceeded. It is interpreted that heating in the presence of water and additional anchor ions allows breakage of the anchored sulfonated groups away from the organic portion of the cation-exchange resin: A water molecule replaces the water molecule which was removed during manufacture of the sulfonated resin, giving back a sulfate; also the hydrogen which had been lost in manufacture returns to the resin. These actions leave BaSO.sub.4 and, locally, the original polystyrene resin. The material that vaporized was near totally condensible at room temperature--very little noncondensible material collected in a ballast vacuum chamber. The condensed vapors were liquid at room temperature, but, after weeks of standing, sometimes show some solid formation due to limited repolymerization. Unlike ion-exchange resin decompositions with NaOH-KOH anchor (bonding) material, which formed some charry residue, as discussed in Case 2, the Ba(OH).sub.2 anchor material did not yield clear evidence of any carbonaceous residue. Apparently the barium hydroxide provides catalysis for depolymerization of ion-exchange resin that NaOH-KOH does not give. The resin depolymerization gives the vapor, and the organic material is largely decomposed. Apparently, even the cross-linked material is decomposed more effectively than in Case 2. The radioactive barium sulfate synthetic barite has not appeared to be wet when the reaction zone is viewed in a borosilicate glass container. Apparently, vaporization largely keeps up with depolymerization. The barite is as crystals which are ghosts of the original ion-exchange resin beads; they are not dusty as they were prepared. The uranium turned black, coloring the barite, but, as noted, there was no obvious carbonaceous deposit. The final location of the decaying atoms was all with the unvaporized residue, as well as was detectable with the Eberline beta-gamma counter used. Experimental Case 2 A typical case with the earlier preferred embodiment using NaOH-KOH bonding material proceeded as follows: Wet cation-exchange resin, U.sup.++++, and solid NaOH-KOH mixtures were heated in a sealable borosilicate glass operated either with vapor boiling or with sweeping argon gas. Heating drove off much of the water as vapor, leaving melted hydroxide mixtures with some extra water in solution. A second liquid phase formed from the solid radioactive ion-exchange resin, and vaporization started. Unlike the condensate with barium hydroxide, as temperatures rose, the heated organic solutions changed both boiling temperatures and colors. Finally, at temperatures approaching 500.degree. C., charry residues remained with the inorganic material, but most of the resin had vaporized. The separation of radioactive from nonradioactive material was again good, with the radioactive material being with the hydroxides, sulfates, and sulfites, which are not as advantageous for permanent radioactive disposal as barite.
	summary
045405123	description	The radioactive stream containing boric acid which is treated in the process of this invention is an aqueous stream which typically comes from the primary loop of pressurized water reactors. That stream may consist of about 10 to 1,000 ppm of boron, usually as boric acid, and may also contain lithium, calcium, magnesium, aluminum, suspended solids such as metal oxides, and radionuclides such as cobalt and various fission products. The stream is usually heated to evaporate some of the water and concentrate the boric acid to about 4 to about 12%. (All percentages herein are by weight unless otherwise indicated.) In the first step of the process of this invention, the aqueous stream containing the boric acid and radionuclides is evaporated to dryness. This can be accomplished in a spray dryer or other type of evaporator, but it is preferably done in a scraped film evaporator because that is one of the few types of evaporators that will take the stream to complete dryness. The drying should be done at temperatures less than 130.degree. F. to avoid volatilizing the boric acid. Preferably, the evaporation should be done under vacuum at temperatures less than 100.degree. F., which can be easily accomplished in a scraped film evaporator. In the second step of this invention, the solids from the first step are mixed with an alcohol to form a boron alkoxide. The alcohol reacts with the boric acid to form the corresponding boron alkoxide and water according to the equation: EQU 3ROH+H.sub.3 BO.sub.3 .fwdarw.B(OR).sub.3 +3H.sub.2 O, where R is alkyl. The alcohol, in addition to reacting with the boric acid to form a boron alkoxide, also forms an azeotrope with the boron alkoxide. Sufficient alcohol should be added to react with all of the boric acid which is present and form the azeotrope. That is, the amount of alcohol should be at least 1.4 times the stoichiometric reaction amount, and up to about 100 mole % in excess of stoichiometric may be used to ensure complete reaction. The alcohol may be methanol, ethanol, propanol, isopropanol, or a mixture thereof, but methanol is preferred as boron methoxide is the most stable boron alkoxide and it forms the lowest boiling point azeotrope, which reduces its chances of decomposition. If ethanol, propanol, or isopropanol is used, it may be necessary to add benzene or another compound which forms an azeotrope with water in order to remove the water as it is formed. In the next step of this invention, the stream is heated to evaporate the boron alkoxide and the alcohol. If methanol is used, they will evaporate together as a boron alkoxide-alcohol azeotrope. If ethanol, propanol, or isopropanol is used, an alcohol-water azeotrope will evaporate first, followed by the boron alkoxide. The methanol azeotrope is 27% methanol and 73% trimethylborate and the temperature of a stream will be 54.degree. C., the temperature at which the azeotrope evaporates. Evaporation is preferably performed by using steam around the jacket of the reactor. Steam at about 212.degree. to 338.degree. F. in an amount of about 21/2 to 3 pounds of steam per pound of boric acid is suitable if 1.4 times the stoichiometric amount of methanol is used. The material remaining in the bottom of the reactor ("crud") may then be solidified in cement, glass, or other materials, according to processes well known in the art. A relatively new way of solidifying this material is to add alkoxide glass formers which are then partially hydrolyzed and heated to form a polymeric glass. (See U.S. Pat. Nos. 4,376,070; 4,377,507; 4,422,965; and 4,430,257), herein incorporated by reference. In the fourth step of the process of this invention, the vaporized boron alkoxide is mixed with water to reform boric acid according to the equation: EQU B(OR).sub.3 +3H.sub.2 O.fwdarw.3ROH+H.sub.3 BO.sub.3. The amount of water used should be at least stoichiometric, but excess water will not be a disadvantage as the boric acid is mixed with water for reuse in the reactor anyway. Recycled boric acid is typically about a 4% solution. In the fifth step of the process of this invention, the alcohol is separated from the boric acid by evaporation of the alcohol to leave a relatively pure boric acid aqueous solution. This is accomplished by heating the stream at the boiling point of alcohol, which is 64.7.degree. C. for methanol, until all of the alcohol has been removed. Again, a scrapped film evaporator is the preferred apparatus, though other evaporators can also be used. The remaining boric acid can then be recycled into the reactor, if desired. The process of this invention can be conducted continuously, in batch, or a mixture of continuous and batch; batch is preferred as it is easier to control. The following examples more fully illustrate the process of this invention. EXAMPLE 1 FIG. 1 is a block diagram which shows the process of this invention using methanol to form boron methoxide, and using alkoxide glass formers to solidify the solids. In FIG. 1, an aqueous solution containing 12% boric acid and 1% crud passes through line 1 to scraped film evaporator 2, which is heated with steam in line 3 producing clean water in line 4. The solids pass through line 5 to boric acid reactor 6. Methanol enters the reactor in line 7 and steam in line 8 heats the reactor. The azeotrope leaves the reactor in line 9 and the crud leaves in line 10. The azeotrope passes to scraped film evaporator 11 and is mixed with water in line 12. Heat from steam in line 13 evaporates methanol in line 7, leaving clean boric acid in line 14. Meanwhile, the crud in line 10 is mixed with alkoxide glass formers in line 15 in mixer 16. Steam heat from line 17 evaporates the methanol in line 18. The solids then move through line 19 to melter 20, where electrical power is applied (line 21), producing crud solidified in glass in line 22. The following table gives the flow rate, temperature, and material balance in the various lines shown in FIG. 1. __________________________________________________________________________ Stream Number 1 4 5 3 9 10 8 18 17 15 7 19 22 __________________________________________________________________________ Flow Rate (lb/hr) 500 427.8 72.2 427.8 147.6 64.6 174 80.1 79 32.2 140 16.7 16.7 Temperature (.degree.F.) 70 120 120 338 158 158 338 212 338 70 70 212 2022 Component (w/o) H.sub.2 O 87.0 100 10.0 100 0 92.2 100 75.7 100 1.21 0 0 0 H.sub.3 BO.sub.3 12.0 0 83.1 0 0 0 0 0 0 0 0 17.5 17.5 Crud 1.0 0 6.9 0 0 7.78 0 0 0 0 0 30.0 30.0 CH.sub.3 OH 0 0 0 0 31.7 0 0 24.3 0 54.5 100 0 0 SiO.sub.2 0 0 0 0 0 0 0 0 0 25.6.sup.2 0 49.3 49.3 Na.sub.2 O 0 0 0 0 0 0 0 0 0 3.39.sup.3 0 6.53 6.53 Al.sub.2 O.sub.3 0 0 0 0 0 0 0 0 0 3.51.sup.4 0 4.43 4.43 B(OCH.sub.3).sub.3 0 0 0 0 68.3 0 0 0 0 11.7 0 0 0 __________________________________________________________________________ Stream 21 -- Electrical Power -- 20 Stream 14 -- 4 w/o H.sub.3 BO.sub.3 in H.sub.2 O -- 1500 Stream 13 -- Steam -- 140 lb/hr Stream 12 -- H.sub.2 O -- 1492.4 Final Volume Reduction -- 105 Final Weight Reduction -- 30 .sup.1 As Al(OH.sub.3) .sup.2 As Si(OCH.sub.3).sub.4 .sup.3 As NaOH .sup.4 As B.sub.2 O.sub.3 EXAMPLE 3 FIG. 2 illustrates a modification of the process shown in FIG. 1 where the solids are encapsulated in cement instead of in glass forming alkoxides. In FIG. 2, crud in line 10 passes instead to cement encapsulation system 24 where water and cement from line 25 are added. The product, crud solidified in cement, leaves by line 26. The following table gives the flow rate, temperature, and material balance for the lines shown in FIG. 2. ______________________________________ Stream Number 10* 25 25 Flow Rate (lb/hr) 19 31.6 50.6 Temperature (.degree.F.) 158 70 70 H.sub.2 O (%) 92.22 0 27.6 Crud (Bottoms) (%) 26.3 0 9.9 B(OCH.sub.3).sub.3 (%) 0 0 0 Cement (%) 0 100 62.5 ______________________________________ *Process #2 removes all H.sub.2 O until the only product is B.sub.2 O.sub.3 Final Weight Reduction 10.1 Final Volume Reduction 26.6
051376814	summary	BACKGROUND OF THE INVENTION 1. Field Of The Invention This invention relates to an improvement in steam power plants for the generation of electricity, and in particular to reducing feed water requirements and condenser complexity by mechanically compressing power turbine exhaust steam to be recycled to the turbine. 2. Description of the Prior Art Although there are many variations of controls and operating elements in steam power plants, the basic elements include a thermal steam source, steam turbines, generators, and one or more condensers to reclaim the turbine exhaust as hot feed water. Various pressure controls and the recycling of dry turbine exhaust steam to super heater tubes are well known variations to the basic power plant to control operating conditions and increase thermal efficiency. Re-compression of turbine exhaust steam by mechanical means to be recycled to the power turbine has not been previously been included in such systems, as the mechanical energy used in compression cannot be entirely extracted as work energy by the turbine, resulting in a net energy loss. However, particularily with nuclear fueled power plants, the large quantities of cooling water and boiler feed water required are, in many locations, more critical limitations than thermal efficiency. Consequently, in such locations it may be advantageous to sacrifice optimum thermal efficiency in order to reduce water requirements. An objective of the present invention is to reduce feed water requirements in the steam loop, and by so doing, also reduce the size and complexity of the condensers. Although mechanically compressed turbine exhaust steam has not previously been used for this purpose, it has been known to mechanically compress steam engine exhaust for auxiliary purposes, such as described in U.S. Pat. No. 1,066,348 to G. T. Voorhees, which describes mechanically compressed exhaust steam utilized for auxiliary heating systems, such as moisture absorption generators. Similarly, U.S. Pat. No. 4,239,603 describes the use of a turbo compressor heat engine to re-pressure a portion of the process vapor from a steam ejection vacuum cooling or drying system, thereby reducing the fuel costs associated with condensing the vapor and regenerating the steam in a boiler. Gas turbine power plants typically extract a portion of turbine power to compress the inlet air prior to combustion, as the work extracted for compression is more than recouped by the increased combustion efficiency. It is also known in gas turbine technology, particularly turbo-prop and turbofan aircraft engines, to use a coaxial dual shaft for the turbine-compressor drive linkage, to accommodate different rotating speeds for the high and low pressure extraction/compression stages. It is an objective of this invention to utilize a dual shaft compresser-turbine in a steam power plant to re-compress the turbine exhaust steam, and to recycle it to the turbine inlet in order to reduce the amount of condensation and steam generation required in the system cycle, thereby sacrificing some thermal efficiency in order to decrease feed water requirements and condenser size and complexity. SUMMARY OF THE INVENTION In the apparatus of the present invention, a steam power plant includes dual shaft turbine-compressor as the power extraction turbine. The turbine exhaust steam is passed through a dryer and then introduced to the compressor section, where its pressure is raised by mechanical compression, and then is mixed with fresh steam by jet ejectors. The mixed steam is thereafter routed to the turbine inlet. In the method of generating electricity by the present invention, dry high pressure steam is generated by a thermal source, is mixed with re-compressed turbine exhaust steam by jet ejectors, and the mixed steam is input to the turbine section of a coaxial dual shaft turbine-compressor. Thermal energy is extracted from the mixed steam by the turbine section and converted into mechanical energy. A portion of the mechanical energy is used to drive an electrical generator and the remaining mechanical energy is used to re-compress turbine exhaust steam. The turbine exhaust steam is passed through a dryer to eliminate moisture, and the dried exhaust steam is routed to the compressor section. The compressor mechanically raises steam pressure, and such re-compressed steam is routed to the jet ejectors to be mixed with fresh steam.
	claims	1. A composition comprising a halide salt matrix having dispersed therein nanoparticles comprising elemental carbon in the absence of water and surfactants, wherein said halide salt is an alkali halide salt, alkaline earth halide salt, or combination thereof, and said halide is fluoride or chloride. 2. The composition of claim 1, wherein said halide salt contains at least one alkali halide salt selected from the group consisting of lithium fluoride, sodium fluoride, potassium fluoride, rubidium fluoride, sodium chloride, potassium chloride, rubidium chloride, and eutectic mixtures thereof. 3. The composition of claim 1, wherein said nanoparticles comprising elemental carbon are selected from the group consisting of hollow carbon nanospheres, exfoliated graphite nanoplatelets, carbon nanotubes, spherical fullerenes, carbon black, carbon nanodiamonds, carbon onions, carbon nanofibers, graphene oxide nanoparticles, reduced graphene oxide nanoparticles, and metal carbide nanoparticles. 4. The composition of claim 1, wherein said composition further comprises nanoparticles comprising a fissile material dispersed within said composition. 5. The composition of claim 4, wherein said fissile material comprises uranium, thorium, or plutonium. 6. The composition of claim 1, wherein said nanoparticles comprising elemental carbon contain a core of a metal encapsulated by a shell of carbon. 7. The composition of claim 6, wherein said metal is a fissile material. 8. The composition of claim 7, wherein said fissile material comprises uranium, thorium, or plutonium. 9. A molten salt reactor comprising (i) a reactor core in which fissile material is housed and (ii) a coolant loop containing a molten salt composition, wherein said coolant loop is within sufficient proximity to said reactor core so as to remove heat from said reactor core and maintain said reactor core within a safe operating temperature, and wherein said molten salt composition comprises a halide salt matrix having dispersed therein nanoparticles comprising elemental carbon in the absence of water and surfactants, wherein said halide salt is an alkali halide salt, alkaline earth halide salt, or combination thereof, and said halide is fluoride or chloride. 10. The molten salt reactor of claim 9, wherein said coolant loop is a primary coolant loop. 11. The molten salt reactor of claim 10, wherein said primary coolant loop is within sufficient proximity to a secondary coolant loop so as to permit said secondary coolant loop to remove heat from said primary coolant loop. 12. The molten salt reactor of claim 11, wherein said secondary coolant loop contains water or supercritical carbon dioxide. 13. The molten salt reactor of claim 9, wherein walls of said coolant loop in contact with the molten salt are constructed of stainless steel. 14. The molten salt reactor of claim 9, wherein said halide salt contains at least one alkali halide salt selected from the group consisting of lithium fluoride, sodium fluoride, potassium fluoride, rubidium fluoride, sodium chloride, potassium chloride, rubidium chloride, and eutectic mixtures thereof. 15. The molten salt reactor of claim 9, wherein said nanoparticles comprising elemental carbon are selected from the group consisting of exfoliated graphite nanoplatelets, carbon nanotubes, spherical fullerenes, carbon black, carbon nanodiamonds, carbon onions, carbon nanofibers, graphene oxide nanoparticles, reduced graphene oxide nanoparticles, and metal carbide nanoparticles. 16. The molten salt reactor of claim 9, wherein said composition further comprises nanoparticles comprising a fissile material dispersed within said composition. 17. The molten salt reactor of claim 16, wherein said fissile material comprises uranium, thorium, or plutonium. 18. The molten salt reactor of claim 9, wherein said nanoparticles comprising elemental carbon contain a core of a metal encapsulated by a shell of carbon. 19. The molten salt reactor of claim 18, wherein said metal is a fissile material. 20. The molten salt reactor of claim 19, wherein said fissile material comprises uranium, thorium, or plutonium. 21. The composition of claim 1, wherein said nanoparticles comprising elemental carbon are hollow carbon nanospheres. 22. The composition of claim 1, wherein said halide salt matrix contains a combination of an alkali halide salt and alkaline earth halide salt. 23. The composition of claim 1, wherein said composition excludes carbon nanotubes.
	description	This application is a continuation of international application no. PCT/EP2009/056863 filed on Jun. 4, 2009, which claims priority from U.S. Provisional application No. 61/058,596 filed on Jun. 4, 2008 and U.S. Provisional application No. 61/179,761 filed on 20 May 2009. All of these applications are hereby incorporated herein by reference in their entirety. 1. Field of the Invention The present invention relates to a method of exposing a target by means of a plurality of beamlets, in particular by means of a charged particle multi-beamlet system, and to a computer readable medium for performing the methods when executed by a processor. 2. Description of the Related Art Systems using a black and white writing strategy, i.e. an “on” and “off” writing strategy, are widely known in the art. They may use, for example, laser beams or charge particle beams, and may feature the use of direct writing means in maskless systems. By modulating the beam (or beams in multi-beam systems), individual grid cells in a rasterized virtual grid may be exposed or not exposed to write the desired pattern on to the target. Such beams are characterized by a so-called beam effect in the target surface, which is often described by a point spread function. The point spread function generally has a Gaussian distribution. The beam size is generally defined as the size of the distribution in which 50% of the beam energy is present. A particular kind of charged particle beam based lithographic system is known from U.S. Pat. No. 6,897,458 assigned to the owner of the present invention, and involves a massive plurality of charged particle beamlets generated in a charged particle beam column for exposing a target. The charged particle beamlets are scanned over the target while being modulated. Additionally, the target may be capable of moving relative to the beams, for example in a direction transverse to the scanning direction of the beams. The modulation of the beamlets is performed on the basis of pattern data provided to the lithographic system. In the particular system described, the modulation is performed by blanking or blocking beamlets to effectively switch the beamlets on and off. Exposing a target using this type of lithography system is achieved by the combination of relative movement of the target and modulation (e.g. timed “on” and “off” switching or blanking) of each charged particle beamlet by the blanker optics. A known method to expose a substrate with beamlets is a raster scan method. In order to accurately expose the target with an exposure pattern, the pattern data is rasterized. The target is positioned on a motor driven stage that is moved in a continuous motion. As the stage is moved, the beam is scanned in a direction substantially perpendicular to the stage motion. By supplying the rasterized pattern data to the system, timed so that the beamlets are modulated in synchronism with the beamlet deflection and stage motion, the pattern represented by the pattern data can be transposed as an exposure pattern onto the surface of the target. The rasterized pattern data corresponds to an exposure pattern on a virtual raster cell grid on the surface of the target. Existing charged particle beam technology is suitable for lithography systems for relatively course patterning of images, for example to achieve critical dimensions of 90 nm and higher. However, a growing need exists for improved performance. It is desired to achieve considerably smaller critical dimensions, for example 22 nm, while maintaining sufficient wafer throughput, e.g. between 10 and 60 wafers per hour. The total current of the beamlets required to achieve a particular throughput, e.g. in lithography defined as a particular number of wafers exposed per hour, depends on the required dose, the area of the wafer, and the overhead time, i.e. the time to move a new wafer into position for exposure. Among others, the required dose in these shot noise limited systems depends on the required feature size and uniformity, and beam energy. For electron beam systems, the maximum single beam current is determined by the spot size. To obtain a good CD uniformity, the required spot size will limit the single beam current to much less than the current required to obtain a high throughput. Thus a large number of beamlets is required (typically more than 10,000 for a throughput of 10 wafers per hour). As the total current through one lens is limited by Coulomb interactions between electrons, the number of lenses in a high throughput system also needs to be large. Increasing the current in the system, results in an increase of the total current on the target. At the same time, however, in order to maintain performance, the number of electrons impinging on the target surface per square critical dimension should be maintained constant. However, designing a system to generate beamlets having a smaller spot size, considerably reduces the charged particle current that may be applied to the target by each beamlet. Irrespective of the brightness of the charged particle source used, the preceding requirements imply a considerably more than linear increase in the number of beamlets in a charged particle multi-beam system compared to the reduction in critical dimension at the same wafer throughput. In order to obtain a high throughput using such a multi-beam system, a beamlet writing strategy is required to expose the desired pattern as efficiently as possible. The relative size of the beamlets and the lenses that focus the beamlets compared to the desired size of the virtual grid limits the possible arrangements of beamlets, so that the beamlets are separated by a sufficiently greater distance than the diameter of the lenses. It is desired that the writing strategy does not require overly complex movement of the beamlets and the target. In this regard, a system providing for uniform deflection of all the beamlets rather than individual control of each beamlet's deflection is desired, particularly for a system having a very large number of beamlets. As a consequence, a simple raster scan of the beamlets as a group may be undesirable due to the incomplete exposure of the areas at the end of each scan of the group of beamlets. The larger the number of beamlets, the greater the effect of these areas of incomplete exposure will have on efficient utilization of the system. It is also desired that the writing strategy reduces the effects of unintended variations between the individual beamlets in a multi-beamlet system. It is therefore desirable to use a method of exposing a target by means of a plurality of beamlets with an improved performance. Embodiments of the invention relate to a method of exposing a target by means of a plurality of beamlets. Additionally, embodiments of the invention relate to a charged particle multi-beamlet system for exposing a target using a plurality of beamlets. In an embodiment, the invention provides a method of exposing a target by means of a plurality of beamlets, the method comprising: providing a plurality of beamlets, the beamlets being arranged in an array; providing a target to be exposed; creating relative movement in a first direction between the plurality of beamlets and the target; moving the plurality of beamlets in a second direction such that each beamlet exposes a plurality of parallel scan lines on the target;wherein the relative movement in the first direction and the movement of the plurality of beamlets in the second direction are such that the distance between adjacent parallel scan lines exposed by the plurality of beamlets is smaller than a projection pitch in the first direction between beamlets of the plurality of beamlets in the array. In an embodiment, the invention provides a charged particle multi-beamlet system for exposing a target using a plurality of beamlets, the system comprising: a beamlet pattern generator for providing an exposure pattern formed by a plurality of beamlets, the plurality of beamlets being arranged in groups of beamlets; an array of projection lens systems for projecting the groups of beamlets on to the surface of the target, each projection lens system corresponding with a group of beamlets; a deflector array for deflecting a group of beamlets in a second direction, the deflector array comprising a plurality of deflectors, each deflector arranged to deflect a corresponding group of beamlets; a substrate support member for supporting the target to be exposed; a control unit arranged to coordinate relative movement between the substrate support member and the plurality of beamlets in a first direction and movement of the group of beamlets in the second direction such that the distance between adjacent scan lines exposed by the plurality of beamlets is smaller than a projection pitch in the first direction between beamlets of the plurality of beamlets in the array.Exposing a target using aforementioned method or system enables full coverage of an area to be exposed by a plurality of beamlets. Additionally, the transmission of charged particles through the system is relatively high. The distance between adjacent scan lines exposed by the plurality of beamlets may equal the projection pitch divided by a positive integer larger than one. The positive integer may equal a factor of the number of beamlets minus one. In this embodiment, full coverage of the area to be exposed is achieved in a very efficient way. In an embodiment, the invention provides a method of exposing a target by means of a plurality of beamlets, the method comprising: providing a plurality of beamlets, the beamlets being arranged in an array; providing a target to be exposed; creating relative movement in a first direction between the plurality of beamlets and the target; moving the plurality of beamlets in a second direction in a plurality of scans such that each beamlet exposes a plurality of parallel scan lines on the target;wherein the relative movement in the first direction and the movement of the plurality of beamlets in the second direction are such that the distance between subsequent scan lines exposed by the same beamlet within the array of beamlets is smaller than the projected size of the array in the first direction, so that scan lines of one or more beamlets from a second scan are interleaved with scan lines of one or more beamlets from a first scan. In an embodiment, the invention provides a charged particle multi-beamlet system for exposing a target using a plurality of beamlets, the system comprising: a beamlet pattern generator for providing an exposure pattern formed by a plurality of beamlets, the plurality of beamlets being arranged in groups of beamlets; an array of projection lens systems for projecting the groups of beamlets on to the surface of the target, each projection lens system corresponding with a group of beamlets; a deflector array for deflecting a group of beamlets in a second direction, the deflector array comprising a plurality of deflectors, each deflector arranged to deflect a corresponding group of beamlets; a substrate support member for supporting the target to be exposed; a control unit arranged to coordinate relative movement between the substrate support member and the plurality of beamlets in a first direction and movement of the group of beamlets in the second direction such that the distance between adjacent scan lines exposed by the plurality of beamlets is smaller than a projection pitch Pproj,X in the first direction between beamlets of the plurality of beamlets in the array.Coordination of relative movement in the first direction and movement in the second direction by using aforementioned method or system enables full coverage of the area to be exposed by the plurality of beamlets. Additionally, the transmission of charged particles through the system is relatively high. Finally, stepwise movement in the first direction may be avoided. The distance between subsequent scan lines exposed by the same beamlet within the array of beamlets may equal N F N - 1 ⁢ P proj , X ,wherePproj,X is a projection pitch in the first direction between beamlets of the array, and FN−1 is a factor of (N−1) unequal to one, and N is the number of beamlets in the array. Obtaining aforementioned distance between subsequent scan lines enables full coverage of the area to be exposed in a very efficient manner without while stepwise movement in the first direction is minimized. Aforementioned relative movement in the first direction may have a constant velocity. Furthermore, the movement in the second direction may be a repetitive movement. The repetitive movement may have a constant frequency. In aforementioned methods according to embodiments of the invention, the method may further comprise defining a virtual grid over the target, the grid providing positions of exposing or not exposing the target by respective beamlets, the exposure or non-exposure in dependence of a blanking or a non blanking of each individual beamlet. The virtual grid may comprise comprises a first axis being oriented in line with the first direction of movement, and a second axis being oriented transverse thereto. In embodiments of the invention, the plurality of beamlets are divided in groups, each group of beamlets being arranged in an array, such that the beamlets of the group do not overlap. As a result, the array of beamlets corresponds to an array of locations in the grid. In aforementioned methods and systems, the projection pitch in the first direction between beamlets of the array may be equal to or smaller than a beamlet spot size as projected on the target. Alternatively, the beamlet spot size as projected on the target may be larger than the distance between adjacent scan lines exposed by the array of beamlets. A small size of the projection pitch in the first direction with respect to the beamlet spot size may enhance averaging out of variation between different beamlets. In embodiments of aforementioned methods and systems, the array of beamlets may comprise a number of rows and a number of columns, at least one of the rows and columns being positioned at an angle unequal to 90° with respect to the first direction and the second direction. The number of rows may be equal to the number of columns. Finally, in an embodiment, the invention provides a computer readable medium for performing, when executed by a processor, an embodiment of aforementioned methods of exposing a target by means of a plurality of beamlets in a charged particle multi-beamlet system. The following is a description of various embodiments of the invention, given by way of example only and with reference to the drawings. FIG. 1 shows a simplified schematic drawing of an embodiment of a charged particle multi-beamlet lithography system based upon an electron beam optical system without a common cross-over of all the electron beamlets. Such lithography systems are described for example in U.S. Pat. Nos. 6,897,458 and 6,958,804 and 7,019,908 and 7,084,414 and 7,129,502, U.S. patent application publication nos. 2007/0064213 and 2008/073588, and co-pending U.S. patent applications Ser. Nos. 61/031,573 and 61/045,243 and 61/055,839, which are all assigned to the owner of the present invention and are all hereby incorporated by reference in their entirety. In the embodiment shown in FIG. 1, the lithography system comprises an electron source 1 for producing a homogeneous, expanding electron beam 20. Beam energy is preferably maintained relatively low in the range of about 1 to 10 keV. The electron beam 20 from the electron source 1 passes a collimator lens 3 to produce a collimated electron beam 21, which impinges on an aperture array 4, which blocks part of the beam and allows a plurality of beamlets 22 to pass through the aperture array 4. The system generates a large number of beamlets 22, e.g. about 10,000 to 1,000,000 beamlets. The plurality of electron beamlets 22 pass through a condenser lens array 5 which focuses each of the electron beamlets 22 in the plane of a beamlet blanker array 6. This beamlet blanker array 6 comprises a plurality of blankers which are each capable of deflecting one or more of the electron beamlets 22. Subsequently, the electron beamlets 22 enter the end module 7. The end module 7 is preferably constructed as an insertable, replaceable unit which comprises various components. In this embodiment, the end module comprises a beam stop array 8, a beam deflector array 9, and a projection lens arrangement 10. The end module 7 will, amongst other functions, provide a demagnification of about 25 to 500 times. Additionally, the end module may be arranged to deflect the beamlets 22. After leaving the end module 7, the beamlets 22 impinge on a surface of a target 11 positioned at a target plane. For lithography applications, the target usually comprises a wafer provided with a charged-particle sensitive layer or resist layer. In the end module 7, the electron beamlets 22 first pass beam stop array 8. This beam stop array 8 largely determines the opening angle of the beamlets. In an embodiment, the apertures in beam stop array 8 are round, resulting in beamlets with a generally uniform opening angle. The passages of the beam stop array 8 may be aligned with the elements of the beamlet blanker array 6. The beamlet blanker array 6 and beam stop array 8 then operate together to block or let pass the beamlets 22 in the following way. If beamlet blanker array 6 deflects a beamlet, it will not pass through the corresponding aperture in beam stop array 8. Instead, it will be blocked by the substrate of beam stop array 8. However, if beamlet blanker array 6 does not deflect a beamlet, it will pass through the corresponding aperture in beam stop array 8, and will then be projected as a spot on the surface of target 11. In this way the individual beamlets 22 may be effectively switched on and off. Next, the beamlets 22 pass through a beam deflector array 9 which provides for deflection of each beamlet 22 in a direction substantially perpendicular to the direction of the undeflected beamlets 22. Next, the beamlets 22 pass through projection lens arrangement 10 and are projected onto a target 11 to be exposed, typically a wafer, located in a target plane. FIG. 2 shows an embodiment of end module 7 in more detail, showing the beam stop array 8, the deflection array 9, and the projection lens arrangement 10, projecting an electron beamlet onto a target 11. The beamlets 22 are projected onto target 11, preferably resulting in a geometric spot size of about 10 to 30 nanometers in diameter. The projection lens arrangement 10 in such a design may provide a demagnification of about 100 to 500 times. In the embodiment of FIG. 2, a central part of a beamlet 22 first passes through beam stop array 8 (assuming it has not been deflected by beamlet blanker array 6). Then, the beamlet passes through a deflector or set of deflectors arranged in a sequence forming a deflection system, of beam deflector array 9. The beamlet 22 subsequently passes through an electro-optical system of projection lens arrangement 10 and finally impinges on a target 11 in the target plane. The projection lens arrangement 10 in the embodiment of the end module shown in FIG. 2, has three plates 12, 13 and 14 arranged in sequence, used to form an array of electrostatic lenses. The plates 12, 13, and 14 preferably comprise plates or substrates with apertures formed in them. The apertures are preferably formed as round holes though the plate, although other shapes can also be used. In an embodiment, the plates are formed of silicon or other semiconductor processed using process steps well-known in the semiconductor chip industry. The apertures can be conveniently formed in the plates using lithography and etching techniques known in the art. The plates may be coated with an electrically conductive coating to form electrodes. A metal with a conductive native oxide may be used for the electrode, such as molybdenum, deposited onto the plate using techniques well known in the art. An electrical voltage is applied to each electrode to control the shape of the electrostatic lenses formed at the location of each aperture. Each electrode is controlled by a single control voltage for the complete array. Thus, in the embodiment shown in FIG. 2 with three electrodes, there will be only three voltages for all lenses. The three control voltages V1, V2, and V3 create an array of uniform electrostatic lenses which focus and demagnify the large number of electron beamlets 22. The characteristics of the electrostatic lenses are controlled by the three control voltages, so that the amount of focusing and demagnification of all of the beamlets can be controlled by controlling these three voltages. In this way, a single common control signal can be used to control a whole array of electrostatic lenses for demagnifying and focusing a very large number of electron beamlets. A common control signal may be provided for each plate or as a voltage difference between two or more plates. The number of plates used in different projection lens arrangements may vary, and the number of common control signals may also vary. The voltages V1, V2, and V3 may be set so that the difference in voltage between the second and third plates (13 and 14) is greater than the difference in voltage between first and second plates (12 and 13). This results in stronger lenses being formed between plates 13 and 14 so that the effective lens plane of each projection lens system is located between plates 13 and 14, as indicated in FIG. 2 by the curved hatched lines between plates 13 and 14 in the lens opening. This places the effective lens plane closer to the target and enables the projection lens systems to have a shorter focal length. It is further noted that while, for simplicity, the beamlet in FIG. 2 is shown focused as from the deflector 9, a more accurate representation of the focusing of beamlet 21, e.g. as in a traced ray illustration, would indicate that the actual lens plane of lens system 10 is between plates 13 and 14. FIG. 2 also illustrates deflection of a beamlet 21 by deflection array 9 in the Y-direction, illustrated in FIG. 2 as a deflection of the beamlet from left to right. In the embodiment of FIG. 2, an aperture in deflection array 9 is shown for one or more beamlets to pass through, and electrodes are provided on opposite sides of the aperture, the electrodes provided with a voltage +V and −V. Providing a potential difference over the electrodes causes a deflection of the beamlet or beamlets passing though the aperture. Dynamically changing the voltages (or the sign of the voltages) will allow the beamlet(s) to be swept in a scanning fashion. Similarly, deflection may also be performed in a different direction perpendicular to the direction of propagation of the beamlets, e.g. in a direction into and out of the paper. The total current of the beamlets required to achieve a particular throughput in a lithography system, i.e. a particular number of wafers exposed per hour, depends on the required dose, the area of the wafer, and the overhead time, i.e. the time to move a new wafer into position for exposure. Among others, the required dose in these shot noise limited systems depends on the required feature size and uniformity, and beam energy. To obtain a certain feature size, i.e. critical dimension (CD), in resist using electron beam lithography, a certain resolution is required. This resolution is determined by three contributions: beam size, resist scattering of electrons, and secondary electrons mean free path combined with acid diffusion. These three contributions add up in a quadratic relation to determine a total spot size, i.e. a beam size projected on the target to be exposed. Of these three contributions the beam size and the scattering depend on the acceleration voltage. To resolve a feature in the resist the total spot size should be of the same order of magnitude as the CD. Not only the CD but also the CD uniformity is important for practical applications. For electron beam systems, the maximum single beam current is determined by the spot size. To obtain a good CD uniformity, the required spot size will limit the single beam current to much less than the current required to obtain a high throughput. Thus a large number of beamlets is required (typically more than 10,000 for a throughput of 10 wafers per hour). For an electron beam system, the total current through one lens is limited by Coulomb interactions between electrons, so that a limited number of beamlets can be sent through one lens and/or one cross-over point. This consequently means that the number of lenses in a high throughput system also needs to be large. So preferably, a very dense arrangement of a large number of low energy beamlets is achieved, such that the multiple beamlets can be packed into an area comparable in size to the size of a typical wafer exposure field. Consequently, the pitch of the apertures in the plates 12, 13 and 14 of the projection lens is as small as possible to create as many electrostatic lenses as possible in a small area. However, reduction in the pitch for a given bore size of the apertures is limited by manufacturing and structural problems caused when the plate becomes too fragile due to the small distances between the apertures, and by possible aberrations in a lens caused by fringe fields of neighboring lenses. By increasing the current in the system, the total current on the target is also increased to limit development of shot noise. At the same time, however, in order to maintain performance, the number of electrons impinging on the target surface per square critical dimension, i.e. per unit of area of CD squared should be maintained constant. It is remarked at this point that spot size in practice corresponds to, and is, as an easier to conceive term, in the remainder of the text used in stead of “point spread function”. Not only a reduced spot size, but also a reduced point spread function of beamlets is desired to maintain sufficient exposure latitude at an increased current. Sufficient exposure latitude requires a relatively high ratio of peak exposure level on the target from a beamlet compared to base or background level of exposure as normally caused by the peripheral Gaussian parts of neighboring beamlets. Designing a system to generate beamlets having a smaller point spread function, however, considerably reduces the charged particle current that may be applied to the target by each beamlet. Irrespective of the brightness of the charged particle source used, the preceding requirements of reduced spot size, increased current, and reduced point spread function imply a considerably more than linear increase in the number of beamlets in the system compared to the reduction in critical dimension at the same wafer throughput. As mentioned earlier, there is a limit to the number of lenses that may be physically realized within a relatively small area that the projection optics, i.e. end module 7 in FIG. 1, may occupy in practical designs. At reduced CD to be achieved, the number of lenses that can be constructed within these dimensions using known techniques is considerably smaller than the number of beamlets required to achieve the desired wafer throughput. A lithography system which provides a solution for aforementioned problems to achieve reasonable throughputs at reduced CD is discussed in U.S. patent application 61/045,243. An embodiment of the lithography system presented in this application is schematically shown in FIG. 4. Embodiments of the lithography system presented in aforementioned application minimize aberration in the system while allowing a disproportionate increase in the number of beamlets in the system. This principal solution using multiple beamlets per projection lens was found after recognition of the fact that with the deflecting action of deflector array 9 in the end module 7, a virtual point of origin of the deflected beamlet is created, so that a virtual beamlet was imagined. FIGS. 3A and 3B are schematic diagrams of the path of beamlets in the end module 7, to illustrate aforementioned insight underlying using multiple beamlets per projection lens. FIG. 3A shows a system having a single beamlet per projection lens system. A single beamlet 22 passes through an aperture in the beam stop array 8, is deflected by deflector array 9, and focused by a projection lens arrangement 10. The deflected beamlet can be considered as a separate ‘virtual’ beamlet arriving at an inclined angle from a different origin than the actual beamlet 22. For example, when beamlet 22 is deflected to the left, it can be regarded as a virtual beamlet 22V originating from a location to the right of the actual origin of beamlet 22, and similarly when beamlet 22 is deflected to the right, it can be regarded as a virtual beamlet 22V originating from a location to the left of the actual origin of beamlet 22. FIG. 3B shows a system having three beamlets per projection lens system, each beamlet originating from a separate point and passing though a projection lens system at different angles. Hence, in accordance with an insight of the present invention, virtual beamlets 22V in FIG. 3A have been replaced by real beamlets 23. The application of multiple real beamlets through a single projection lens system appears to be possible without disturbing the amount of aberration. This is especially the case where all beamlets present in the lithography system are distributed over a multiplicity of projection lens systems. As compared to FIG. 3A, which uses a single deflected beamlet, the net effect is the same, except that the current through the projection lens is three times greater in the system of FIG. 3B. Aperture plate 4 above the deflector array 9 may comprise a multiplicity of apertures 23 as depicted i.e. may have a patterned opening to accommodate the multiple beamlets 22 that correspond to a single lens system associated with on opening in beam stop array 8. Note that in a real design, the beam angle is much smaller than shown in FIGS. 3A and 3B. The relatively large beam angle shown in FIGS. 3A and 3B is merely present to clarify the invention. Because part or all of the plurality of beamlets directed through to a each projection lens system may be blanked at any point in time during operation, a system encompassing aforementioned principle may in this description be referred to as a patterned beamlet system. Such a patterned beamlet system may also be regarded as a multiplicity of miniaturized imaging systems arranged side by side. FIG. 4 illustrates an embodiment of a charged particle multi-beamlet lithography system including grouped beamlets. This embodiment is able to accommodate an increased number of beamlets in the system, permitting increased current at the wafer and/or reduced spot size. The embodiment shown in FIG. 4 is constructed generally as described for the system of FIG. 1, except that the beamlets are arranged in groups so that multiple beamlets may be focused by a single projection lens system. In this embodiment, an aperture array 4A produces beamlets 22 from the collimated beam 21. The beamlets 22 are focused by condenser lens array 5A in the plane of a second aperture array 4B. The second aperture array 4B may be separate from or combined with a beamlet blanker array 6. The beamlets 22 are arranged as groups, and a second condenser lens array 5B focuses each group of beamlets towards a corresponding aperture in beam stop array 8. The beamlet blanker array 6, which may be integrated with the second aperture array 4B or constructed separately, comprises an array of blanker elements with electrodes formed at the location of each aperture in beamlet blanker array 6. Each blanker element operates to deflect a beamlet 22 in a group of beamlets. Although three beamlets per projection lens system is shown in FIG. 4, other numbers of beamlets per projection lens system may also be used, and groups of up to 100 beamlets or more can be directed through each projection lens system. In a preferred embodiment, groups of 49 beamlets in an array of 7 by 7 are deflected through each projection lens system. The apertures in the beam stop array 8, which define the beamlet opening angle, may be relatively small as if they were limiting only a single beamlet. Larger apertures would require a larger deflection path, would be more susceptible to “tail” effects caused by only partial blanking of a blanked beamlet, and would further reduce the limited space available on beam stop array 8 for blanking beamlets. In principle each group of beamlets can be converged, i.e. directed to a single point, i.e. spot, where they intersect and cross-over, either at the relevant aperture of beam stop array 8, or at the effective lens plane of the relevant projection lens system. In practice the convergence will be somewhere between these two points (although not further depicted in the drawings), since converging the beamlets at the beam stop array would create a lens error while converging the beamlets at the effective lens plane of the projection lens would cause a dose error. In this embodiment of a charged particle multi-beamlet lithography system multiple beamlets pass through each projection lens system. The charged particle optics slit does not consist of a regular array of beamlets but of a regular array of groups of beamlets. As discussed in U.S. patent application 61/045,243, at any instant some of the beams in a group may be directed through a corresponding opening in beam stop array 8 and projected onto the target, while other beamlets are deflected an additional amount. This additional deflection causes these beamlets to miss the opening in beam stop array 8 so they are blocked from reaching the target, and are thereby blanked or “switched off” as described previously. Thus, each group of beamlets exposes a pattern determined by the beam blanker array 6, and each group can be considered as a single patterned beamlet. FIG. 5 illustrates a simplified schematic overview of an embodiment of the charged particle multi-beamlet lithography system including beamlets formed from sub-beams. While the system of FIG. 4 provides for multiple beamlets per projection lens system, it also results in a complex system requiring two sets of condenser lens arrays 5A and 5B. Furthermore, manufacturing the condenser lens array 5A is difficult as it comprises relatively small openings at a fairly small pitch. The lithography system of FIG. 5 is less complex, and as a result of a decrease in the number of components, has a reduced column length as compared to the lithography system illustrated in FIG. 4. Due to the decreased column length, a smaller vacuum chamber may be used. Such a vacuum chamber is not only less expensive, but it can also be pumped out in a shorter period of time, i.e. its pumping time is limited. Furthermore, the path length of the beamlets is decreased, resulting in a decrease of misalignment due to beamlet drift. The system of FIG. 5 includes an aperture array 4C to produce larger sub-beams 25. The sub-beams pass through a condenser lens array 5 focusing the sub-beams approximately in the plane of beam stop array 8 and towards a corresponding opening in beam stop array 8. In principle each sub-beam can be focused either at the corresponding opening in beam stop array 8, or at the effective lens plane of the corresponding projection lens system. In practice it is preferred to focus the sub-beams somewhere between these two points. This results in the aperture array 4C being imaged onto the target 11. It should be noted that the condenser lens array (in any of the embodiments) may comprise a single condenser lens array or a set of condenser lens arrays, as would be known to a person of skill in the field of electron-optics. The sub-beams 25 are intercepted by aperture array 4D which includes a number of apertures in the path of each sub-beam, producing a group of beamlets 23 from each sub-beam 25. The groups of beamlets, formed from the sub-beams, are also focused approximately in the plane of beam stop array 8 and each group of beamlets is directed towards a corresponding opening in beam stop array 8. These beamlets 23 then pass through beamlet blanker array 6. Blanked beamlets are blocked by beam stop array 8, while unblanked beamlets in each group pass through a corresponding opening in beam stop array 8 and are subsequently projected onto the target by projection lens arrangement 10. In the example shown in FIG. 5, the aperture array 4D produces a group of three beamlets 23 from each sub-beam 25. The group of beamlets, if undeflected by beam blanker array 6, strike the beam stop array 8 at a corresponding opening so that the three beamlets are projected onto the target by the projection lens system 10. In practice, a much larger number of beamlets may be produced for each projection lens system 10. In a practical embodiment, as many as 50 beamlets may be directed through a single projection lens system, and this may be increased to 200 or more. As shown in FIG. 5, the beamlet blanker array 6 may deflect individual beamlets 23 in a group of beamlets at certain times in order to blank them. This is illustrated in FIG. 5 by the left-hand sub-beam 25, in which the middle beamlet 23 has been deflected to a location on the beam stop array 8 near to but not at an opening so that the beamlet is blanked. In the middle sub-beam 25 the right-hand beamlet 23 has been deflected and is blanked, and in the right-hand sub-beam 25 no beamlets are deflected and blanked. As schematically shown in FIGS. 4 and 5, a control unit 18 is arranged to actuate movement of a substrate support member or stage 16 arranged for supporting the target 11 to be exposed, e.g. a wafer. This movement is referred to variously in this description as a movement in a first direction or the X-direction or the mechanical scan direction. The movement in the X-direction can be achieved in a variety of ways: the target moves by mechanical movement of the substrate support member, the rest of the system moves, the beamlets are deflected, or any combination of the above techniques. As mentioned with respect to FIG. 1, the patterned beamlets formed in the charged particle multi-beamlet lithography systems shown in FIGS. 4. and 5 are moved by a deflector array 9. This movement of the beamlets, in this case the deflection of charged particle beamlets by the electrostatic deflector array 9, is performed in a second direction, the second direction being different than the first direction. This second direction is referred to variously in this description as the Y-direction or deflection scan direction. In one embodiment of the invention, the first direction is substantially perpendicular to the second direction. Please note that the movement of the beamlets in the second direction is generally a repetitive movement, i.e. a beamlet is scanned from a starting position over a certain distance until an end position and then moved back again to the starting position. Hence, the movement of the beamlets in the second direction may be performed with a certain frequency. As mentioned previously, for lithography applications, the target usually comprises a wafer provided with a charged-particle sensitive layer or resist layer. The current industry standard is a 300 mm wafer. Wafers are typically divided into fixed size fields with a maximum dimension of 26 mm×33 mm. Each field may be processed to produce multiple integrated circuits (i.e. the layout for multiple chips may be written into a single field) but the ICs do not cross a field border. With a maximum size of 26 mm×33 mm there are 63 fields available on a single standard wafer. Smaller fields are possible and will result in a higher number of fields per wafer. FIG. 6 shows a wafer 26 divided into fields 27, and the direction of writing the fields 28. It is possible to write partial (incomplete) fields, for example by writing full fields into the partial fields and crossing the wafer boundary. In a preferred embodiment of the lithography machine, the machine generates 13,000 sub-beams and each sub-beam is split into 49 beamlets arranged in a 7×7 array, resulting in 637,000 beamlets (i.e. 13,000×49). When this arrangement is applied in the machine of FIG. 5, the aperture array 4C contains 13,000 holes and the aperture array 4D and beamlet blanker array 6 contain 637,000 holes, which form the electron-optical (EO) slit in an area of 26×26 mm (i.e. the size of the complete array of beamlets as projected onto the wafer). Each sub-beam comprising 49 beamlets writes a single stripe in each field in the X-direction. The 13,000 sub-beams arranged over a distance of 26 mm (i.e. the width of the EO slit) result in a stripe of width 2 μm in the Y-direction (substantially perpendicular to the mechanical scan) and a stripe length as long as the field in X-direction (e.g. 33 mm in the example of FIG. 6). The wafer is preferably written on (exposed) by the lithography machine in both a backward and forward X-direction, for example as shown in FIG. 6. The direction of writing in the Y-direction (by the deflector array) is usually in one direction. When the size (height) of a field is chosen to be smaller than the size of the electron-optical (EO) slit (i.e. the size of the complete array of beamlets as projected onto the wafer) (e.g. smaller than the maximum size of 26 mm), then more fields can be placed on the wafer, but not all of the electron beamlets will be used to write on the wafer. The EO slit will need to cross the wafer more often to write all fields and the overall throughput will decrease. Therefore, the EO slit preferably has a size that coincides with the size of a field. When a machine is writing patterns to a field, at some moment the beamlet blanker array enters the next field and starts writing patterns in it, so the machine should be able to write in two fields at the same time. If a field is sufficiently small, the machine should be able to write 3 fields at the same time. A simplified version of a beamlet blanker array 35 is shown in FIGS. 7A and 7B, for a machine having 16 sub-beams arranged in a 4×4 array, and each sub-beam having 9 beamlets arranged in a 3×3 array. For each sub-beam there is a group of apertures 36 with an associated blanker electrode for each beamlet aperture. An expanded view of one group of apertures 37 is shown on the right side of the drawings. In FIG. 7A an arrangement of blanker apertures is shown suited to a parallel projection writing strategy, while in FIG. 7B an arrangement is shown suited to a perpendicular writing strategy. In FIG. 7A the blanker apertures for the beamlets of each sub-beam are grouped closer together. In FIG. 7B the blanker apertures for the beamlets are distributed evenly over the full stripe width 39, i.e. each beamlet is positioned equidistant from the neighboring beamlets (whether from the same sub-beam or a different sub-beam) in a Y-direction substantially perpendicular to the mechanical scan direction. This is possible, but for a low number of apertures the efficiency of this arrangement in terms of the ratio between the beam and beamlet current will be extremely low. A measure for the efficiency is the fill factor, which is the ratio between the total area of the blanker apertures and the area in which the apertures for one sub-beam are grouped. The fill factor is useful for evaluating the efficiency of a particular grid geometry in terms of current in (beam current) and current out (sum of beamlet currents). When the area of the group of beamlet apertures is smaller, the fill factor will increase to more preferred values. A writing strategy that works well for a small number of apertures is the “parallel projection” writing strategy, where (in its simplest form) the individual beamlets are interleaved and write the full stripe width. For example, in the lithography machine discussed above, a number of apertures for which the parallel writing strategy would be suitable would be in the order of 750,000 or smaller. As will be explained later in further detail, a parallel writing strategy refers to a writing strategy in which a direction of the sequence of projected beamlets is parallel to a (mechanical) scan direction of a wafer stage. It is assumed that an ideal grid exists on the wafer and that the electron beamlets can be positioned substantially on the grid coordinates. Assuming that an acceptable exposure pattern is created on the target when the electron beamlets can be positioned within 1/20th of the minimum feature size (the CD), then to write a square having a side equal to the CD, 20 scan lines and thus 20*20=400 grid points are needed. Thus, if the CD equals 45 nm, in this embodiment the positioning is controllable within a range of 2.25 nm. Details with respect to positioning of features by means of gray writing, and with respect to correction for positioning errors by means of timing can be found in U.S. Pat. No. 6,897,458. FIG. 8 shows a beamlet size 30 of a charged particle beamlet projected onto a target surface within a portion of a virtual grid covering an area of a typical desired critical dimension (CD). In order to expose the target with an exposure pattern, the pattern data may be rasterized corresponding to a virtual grid on the surface of the target. In a typical arrangement, the target is moved while the beamlets are scanned in a direction substantially perpendicular to the stage motion. The rasterized pattern data is supplied to the multi-beamlet system, timed so that the beamlets are modulated in synchronism with the beamlet deflection and target motion, so that the pattern represented by the pattern data is transposed as an exposure pattern onto the surface of the target. Although the virtual grid is typically a Cartesian grid, other types of grid are also possible. The size of the grid used may be determined by considering whether an unintentional modulation or misalignment of a single charged particle beamlet would be disturbing to the pattern to be exposed on the target. A tendency exists to choose the grid as small as possible, due to the desire to have a virtually infinite choice in designing line or object width or determining positioning of a feature to be formed on the target. The latter would, in accordance with an insight underlying the present invention, mean an additional possibility for correcting proximity effects at writing. On the other hand, it is desired to have the grid as large as possible, particularly in multi-beam systems having a large number of beamlets, in order to limit the amount of data to be processed and transferred to the system. As a balance between the above described conditions and as a reflection of the state of contemporary technology, one embodiment of a multi-beamlet system discriminates so-called critical dimension cells, typically e.g. 45 nm. These cells may be exposed by charged particle beamlets having a beam size of a corresponding order, e.g. 30 nm. The cells may be divided into a plurality of grid cells, e.g. 20 by 20 grid cells, thus having a small dimension relative to the beam size, e.g. 2.25 nm. In such an arrangement, an accidental blanking or not blanking of a single grid cell would only have a negligible effect. The plurality of grid cells form a virtual grid. The position of the projected beamlet size 30 in FIG. 8 on the virtual grid corresponds to grid cell 31. Since the projected beamlet size 30 of the beamlet in FIG. 6 is much larger than the size of a single grid cell 32, e.g. 2.25 nm (also referred to as the pixel size), edge shapes are largely invisible after developing an exposed pattern. Furthermore, a hatched, i.e. ragged, edge in a rasterized pattern may effectively shift the position of the eventual feature edge, as is in more detail explained in US patent application publication no. 2008/073588. The possible accuracy of placement using the aforementioned technique of ragged edges depends on the size of the ragged edge pixels length compared to the beam size 30. Optimization may result in an accuracy of less than 1/10 of the pixel size. FIG. 9 schematically shows the concept of projecting a group of beamlets 40A, 40B, 40C, 40D onto a plane in a X-direction or a Y-direction. The group of beamlets is arranged in an array as projected on a target surface to be exposed. This concept is used in embodiments of the invention. In FIG. 9, the group of beamlets 40A-40D comprises an array consisting of 2 columns and 2 rows. The distance between the beamlets, also referred to as pitch, is equal to Pb. The beamlets may be moved in the X-direction, e.g. the mechanical scan direction, and the Y-direction, e.g. the deflection scan direction. A combination of movement in these directions may be used in embodiments of the invention. At least one of the rows and columns is positioned at an angle unequal to 90° with respect to the first direction and the second direction. In the example shown in FIG. 9, the columns formed by beamlets 40A, 40B and 40C, 40D are positioned at an angle α with respect to the Y-direction. The beamlets 40A-40D may be projected in the Y-direction onto a plane parallel to the X-direction, resulting in a sequence of beamlet projections 41A-41D (shown as hatched circles) aligned in a direction parallel to the X-direction. The pitch in this projection (i.e. distance between the beamlets when projected onto a plane parallel to the X-direction) is denoted by Pproj,X, and this is referred to hereafter as the projection pitch in the X-direction. The pitch Pproj,X can be calculated by the equation Pproj,X=Pb·sin(α). The grid width Wproj,X for the group of beamlets is also shown. Similarly, beamlets 40A-40D may be projected in the X-direction onto a plane parallel to the Y′-direction, resulting in a sequence of beamlet projections 42A-42D (shown as hatched circles) aligned in a direction parallel to the Y′-direction. The pitch in the projection is denoted by Pproj,Y, and this is referred to hereafter as the projection pitch in the Y′-direction. In this description, several embodiments of methods of exposing a target by means of a plurality of beamlets will be disclosed. The methods can be divided in two categories. A first category, in which the projection pitch Pproj,X, is taken into account, will be referred to as a writing strategy of the parallel type. The expression “parallel” here refers to the direction of the sequence of projected beamlets with respect to the X-direction (e.g. mechanical scan direction). A second category, in which the projection pitch Pproj,Y is taken into account, will be referred to as a writing strategy of the perpendicular type. As mentioned earlier, a target is exposed in a desired pattern by means of a plurality of beamlets by simultaneous movement of the plurality of beamlets relative to the target in a first direction (e.g. a mechanical scan direction) and second direction (e.g. deflection of the plurality of beamlets in a second deflection scan direction). The two directions are not the same. For convenience the drawings show the two directions as being perpendicular. However, in a preferred embodiment the two directions are substantially (but not exactly) perpendicular. This is indicated in the drawings by reference to the Y′-direction. The Y′-direction is slightly tilted with respect to the Y-direction, so that movement of the beamlets in the Y′-direction is coordinated with the simultaneous relative movement of the beamlets and target in the X-direction, resulting in the end of a scan line for a beamlet being at approximately the same position in the X-direction as the beginning of the next scan line for that beamlet. The amount of tilt depends on the amount of movement in the Y-direction during the exposure of a single scan line in the X-direction. The discussion of embodiments of the invention will be illustrated for an array of 2×2 beamlets per projection lens system (i.e. per sub-beam). Although four beamlets per projection lens system are shown, other numbers of beamlets per projection lens system may also be used. Groups of up to 100 beamlets or more can be directed through each projection lens system. Furthermore, although an array is shown in which the number of column equals the number of rows, other array configurations may be used in embodiments of the invention. Hence, an array may comprise n rows×k columns, n and k being integers not equal to each other. Furthermore, although a square array is shown, other arrangement may be used, such as a triangular, hexagonal, or other type of array. In a preferred embodiment the beamlets are charged particle beamlets, preferably electron beamlets, but other types of beamlets suitable for exposing a target may also be used. Finally, note that the drawings do not give an accurate representation of the ratio between dimensions in the X-direction and Y′-direction. In particular, the scan lines in the Y′-direction have been shortened to enhance clarity, especially with respect to effects occurring at the ends of a scan line. FIG. 10 shows a schematic diagram of a scan line of a beamlet. The beam deflector array generates a triangular shape deflection signal for all beamlets in parallel. The deflection signal includes a scan phase (from A to B) and a fly-back phase (from B to C), as shown in the schematic diagram in FIG. 10. During the scan phase, the deflection signal slowly moves the beamlet (when switched on) in the Y′-direction and the beamlet blanker array will switch the beamlet on and off according to the beamlet control signals. After the scan phase the fly-back phase starts. During the fly-back phase, the beamlet is switched off and the deflection signal quickly moves the beamlet to the position where the next scan phase will start. A scan line is the path of a beamlet on the surface of the wafer during the scan phase. As mentioned above, in certain embodiments, without special measures the scan line would not write exactly along the Y-direction on the wafer, but will be slightly skewed with a small X-direction component as well because of the continuous stage movement in the X-direction. This error may be corrected by adding a small X-direction component to the deflection field to match the stage movement. This X-direction component is typically small because the stage movement is slow in comparison to the Y-direction deflection scan speed (a typical x:y relative speed ratio may be 1:1000). However, the effect of this X-direction component is greatly increased in systems with patterned beams (e.g. multiple beamlets for each sub-beam). Firstly, the deflection speed may be reduced in proportion to the number of beamlets per patterned beam (sub-beam). Secondly, due to the inclination of the array of beamlets (as shown, for example, in FIGS. 7A, 7B), a skew in the scan lines on the wafer will result in altering the distance between scan lines made by different beamlets. A large enough skew may result in scan lines overlapping or changing position with respect to each other. A scan line (see FIG. 10 at the right) is divided into three sections: a start overscan section, a pattern section, and an end overscan section. Beamlets are deflected along the Y′-direction across the width of a stripe. The distance in which the beamlets are deflected is typically wider than the actual stripe to be written. Overscan provides space for shifting and scaling the positions where the beamlet writes. Overscan is the single sided surplus. In case of a stripe width of 2 μm and an overscan of 0.5 μm (or 25%) this results in a scan line length of 3 μm. During the overscan sections the beamlets are typically switched off. In FIG. 10 (at the left) a scan line is depicted for the situation where only one beamlet is writing a stripe. The path of the beamlet during a deflection cycle is A-B-C. AB is the scan line movement during the scan phase, while BC is the fly-back during which the beamlet is switched off. The stripe borders are marked D and E. At the right side in FIG. 10 the overscan and pattern sections are identified. The total set of bits of the beamlet control signal received by the beamlet blanker array for switching the beamlet over the scan line is called the scan line bit frame. During the entire scan line the beamlets are controlled by the lithography system. In the overscan section the beamlets will be switched off. In the pattern section the beamlets are switched according to the features required to be written in the wafer field. The bits in the scan line bit frame for both the overscan section and pattern section represent data to be transferred to the beamlet blanker array. The bits/pixels in the overscan section seem to be useless and consuming bandwidth of the data path. However, the bit/pixels in the overscan section can provide room for corrections (such as pattern shift and pattern scaling), provide room for stitching algorithms, and provide room for differences in y-positions of blanker holes for beamlets when writing strategies are used where all beamlets write the full stripe width (parallel projection). Assuming a fixed bit rate for the beamlet control signals which control the beamlets and a certain pixel size, the scan line can be mapped into a fixed length bit frame, referred to as the scan line bit frame. In a preferred embodiment, each sub-beam is split into 49 beamlets for writing one stripe of a field. There are many different writing strategies for writing the stripe. The beamlet writing strategy defines in what way the beams are arranged for writing a stripe. The scheme could be combinations of stacking, interleaving or overlapping. In FIG. 11 several examples are shown as possible interleaving schemes for writing a stripe using four beamlets. The width of one stripe is shown at the right of the drawing. These examples do not show how the beamlets write in real-time, but show which beamlet has written which part of the stripe when the writing has finished. Example A in FIG. 11 shows stacking the beamlets. Every beamlet writes in its own substripe. For this configuration each beamlet only writes a small number of bits before it flies back. The frequency of the deflection signal is high and its amplitude is low. This writing strategy is suited to the case where the beamlets in a group are arranged so that the group width GW (equal to the number of beamlets N in a sub-beam group multiplied by the projection pitch Pproj,X) is equal to the stripe width (perpendicular projection). Perpendicular projection is a family of writing strategies. For the basic form of perpendicular projection all beamlets write small substripes, and the width of the substripe is a fraction of the stripe width. The size of the grid of blanker holes is typically related to the stripe width. In example B, the beamlets are interleaved over the full stripe width. The frequency of the deflection signal is low and its amplitude is large. The writing strategy that fits with the interleaving scan lines is the parallel projection writing strategy. Parallel projection is a family of writing strategies, in which all beamlets write the full stripe width in an interleaved way. Note that the grid of blanker apertures is not necessarily related to the stripe width. Especially for a relatively low number of beamlets in a group, this strategy allows for a smaller group size and an improved fill ratio. Because of the small number of beamlets, the size of the group on the wafer is significant smaller than the stripe because of a reasonable fill factor. For this write strategy (parallel projection) one can calculate a series of pixel sizes that is realized for a particular number of beamlets in a group and a certain beamlet pitch. The pixel size is not an arbitrary value. Extra bits in the scan line bit frame may be added to compensate for the worst case offset between a beamlet blanker aperture and the center of the stripe. Example C is a combination of interleaving and stacking. For example D the successive interleaved layers are overlapping like a brick wall. Compared to example C, this configuration will provide better averaging between beamlets. At the stripe boundary there are beamlets that would write over the stripe boundaries. FIGS. 12A and 12B schematically illustrate methods of exposing a target by means of a plurality of beamlets in which the exposure pattern is achieved by deflection of each beamlet over the full scan width to be covered by an individual projection lens system. For example, for a field to be exposed with a width of 26 mm using a system having 13,000 projection lens systems, an arrangement of the projection lens systems will result in a full scan width of 2 microns. In this example, the target is exposed by using a writing strategy taking into account the projected pitch of the beamlets in the array as projected onto a plane in the X-direction, i.e. Pproj,X. The distance between adjacent scan lines is equal to the projection pitch in the X-direction. In FIG. 12A, a part of the exposed area of a target is shown being exposed by four beamlets along vertical scan lines. An area exposed by a first beamlet, e.g. beamlet 40A in FIG. 9, is schematically depicted by the scan lines with a gray pattern. An area exposed by a second beamlet, e.g. beamlet 40B in FIG. 9, is schematically depicted by the scan lines with a first hatched pattern, i.e. hatching sloping down to the left. An area exposed by a third beamlet, e.g. beamlet 40C in FIG. 9, is schematically depicted by the scan lines with a second hatched pattern, i.e. hatching sloping down to the right. An area exposed by a fourth beamlet, e.g. beamlet 40D in FIG. 9, is schematically depicted by the scan lines with a third hatched pattern, i.e. a cross-hatched pattern. In the example shown in FIG. 12A, the desired grid cell size (i.e. the desired beamlet positioning resolution) is smaller than the projection pitch of the beamlets in the X-direction. It can be readily seen that exposing a target by moving the array of beamlets relative to a target in the X-direction while deflecting the beamlets in a straightforward way in the Y′-direction will not provide full coverage of the target. Hence, in order to provide full coverage a different writing strategy is needed. In FIG. 12B, the area is fully exposed by using an alternative writing strategy, employing a writing strategy of the parallel type. In this method, the relative movement in the X-direction and the movement of the plurality of beamlets in the Y′-direction are such that the distance between adjacent scan lines exposed by the plurality of beamlets equals P proj , X ( F N - 1 ) where FN−1≠1. In this equation, Pproj,X is a projection pitch between beamlets of the plurality of beamlets in the array in a direction parallel to the X-direction; and FN−1 is a factor of (N−1), where N is the number of beamlets in the array. In the embodiment schematically depicted in FIG. 12B, four beamlets are used. Hence, the relative movement between the target and the plurality of beamlets is performed such that the distance between adjacent scan lines exposed by the plurality of beamlets equals ⅓ of Pproj,X. In case the number of beamlets in the array would be 100, e.g. an array of 10×10 beamlets, the distance between adjacent scan lines exposed by the plurality of beamlets according to this method would equal 1/99, 1/33, 1/11, 1/9, or ⅓ of Pproj,X for FN−1 being 99, 33, 11, 9 and 3 respectively. Furthermore, in this embodiment the relative movement between the target and the beamlets is performed in a stepwise manner. Movement in the mechanical scan direction is such that the beamlets may expose three scan lines before reaching an area already exposed by another beamlet. This movement is schematically depicted with respect to the second beamlet by means of a small arrows at the bottom of FIG. 12B. After finishing the third scan line, the beamlets are translated over a larger distance, such that the first beamlet is positioned just after the last scan line exposed by the fourth beamlet. This translation is schematically depicted with respect to the second beamlet by means of a large arrow. After the translation, a new sequence of three scan lines is started. The stepwise movement of this writing strategy may be accomplished by a constant deflection pattern of the beamlets and varying the speed of movement of the target in the X-direction, or a varying deflection pattern of the beamlets and steady movement of the target, or a combination of both. The advantage of the writing strategy used to obtain the exposure as shown in FIG. 12B over the writing strategy used to obtain the exposure as shown in FIG. 12A is that the entire area of the target is covered. However, movement in the mechanical scan direction, i.e. direction of intended mechanical movement of the target is irregular, i.e. requires the deflector 9 of the system to jump over already exposed scan lines in steps of irregular magnitude. Especially with respect to the translation over a large distances, the deflector may require relatively considerable settling time, thereby spending overhead time of the system. Moreover, such system and writing method could be susceptible to positioning errors, hence, necessitating to account for settling times and/or implement positioning control to avoid so-called stitching errors. A disadvantage of the writing strategy is that a certain portion of adjacent scan lines are exposed (scanned) by the same beamlet. As a result, variations in the brightness or strength between individual beamlets may result in undesired variations in the exposure pattern on the target. The averaging effect achieved by the writing strategy of FIG. 8A, in which adjacent scan lines are exposed by all of the beamlets in the array, is thus reduced. This disadvantage may be reduced where the beamlet size is much larger than the width of the adjacent scan lines exposed by the same beamlet. FIGS. 13A-13C schematically illustrate another embodiment of a method of exposing a target. The method enables full exposure of the target without suffering from the disadvantages of the method presented with reference to FIG. 12A, and potential drawbacks of the method presented with reference to FIG. 12B. The method exemplified in FIGS. 13A-13C is a writing strategy of the parallel type. In this method, the relative movement in the first direction and the movement of the plurality of beamlets in the second direction are coordinated so that the distance between subsequent scan lines exposed by the same beamlet within the plurality of beamlets equals N F N - 1 ⁢ P proj , X where FN−1≠1. In FIG. 13A, the result of exposure of a first scan line by each beamlet in a 2×2 array of beamlets is schematically shown. In FIG. 13B, exposure of the target has progressed so that each beamlet has exposed a portion of a second scan line. In FIG. 13C, exposure has progressed further such that an area of full coverage of the target is obtained. In this method, instead of having a relative movement of a single scan line width in the X-direction during exposure of a single scan line, the relative movement in the X-direction during exposure of a single scan line equals N F N - 1 ⁢ P proj , X where FN−1≠1. In FIGS. 13A-13C, four beamlets (in a 2×2 array) have been used to illustrate this method. Hence, the distance between subsequent scanlines exposed by the same beamlet within the plurality of beamlet equals 4 3 ⁢ P proj , X as is schematically shown in FIGS. 13B and 13C. In case the number of beamlets in the array would be 100, e.g. an 10×10 array of beamlets, the distance between subsequent scan lines exposed by the same beamlet within the plurality of beamlets according to this method would equal 1 ⁢ 1 99 ⁢ P proj , X , 3 ⁢ 1 33 ⁢ P proj , X , 9 ⁢ 1 11 ⁢ P proj , X , 11 ⁢ 1 9 ⁢ P proj , X ⁢ ⁢ or ⁢ ⁢ 33 ⁢ 1 3 ⁢ P proj , X for FN−1 being 99, 33, 11, 9 and 3 respectively. The method as used in FIGS. 13A-13C provides full coverage of the target to be exposed along the X-direction for a width equal to L1, i.e. between the dashed-dotted lines in FIG. 13C. The areas outside of the area L1 may be exposed by one or more other projection lens systems of a multi-beamlet system, e.g. the charged particle multi-beamlet system schematically depicted in FIG. 4 or FIG. 5. As can be seen in FIG. 13C, a portion of each scan line falls outside the area of full coverage, labeled L2 and L3. During the time beamlets are deflected outside the full coverage area, they are typically not used (i.e. blanked or blocked), resulting in less efficient utilization of the multi-beamlet system. This effect is lessened when the full coverage length L1 is much larger than the “end area” lengths L2 and L3. The writing strategy of FIG. 13C avoids the stepwise movement of the previous writing strategy. Adjacent scan lines are exposed by different beamlets, providing the advantage of averaging out variations between beamlets when the beamlet size is greater than the scan line width. In the example shown, all four beamlets are scanned over four adjacent scan lines, resulting in a 4× averaging effect. As mentioned earlier, the pattern data for the pattern to be exposed on the target is typically rasterized with respect to a virtual grid. In order to expose the pattern on the target, each beamlet needs to be switched on and off or otherwise modulated in such a way that the sequence of scan lines exposed by the plurality of beamlets form the desired pattern on the target. With respect to the aforementioned methods, a pattern to be exposed by a single projection lens system of a multi-beam system is divided into scan lines over the full width of L1. The scan lines are then assigned to beamlets. Thus, in dependence of the writing strategy used, different scan lines may be assigned to different beamlets. FIG. 16 shows a table of pixel size and grid width in dependence on the number of beamlets per patterned beam N, the array tilt angle α, the projection pitch Pproj,X, and K factor. A large pixel size is desired in order to reduce the amount of control data needed to be generated and transmitted to the lithography machine and to increase throughput of the machine. However, the size of the pixels are limited by the desired CD and the resist properties. In FIG. 16, an optimum pixel size LpixX of 3.5 nm in the X-direction is assumed, and the fourth column from the left shows calculated values of K based on the projection pitch and optimum pixel size (i.e. projection pitch Pproj,X divided by pixel size LpixX of 3.5 nm). The closest value of K which is acceptable given the number of beamlets per patterned beam (as explained above), is shown in the fifth column from the left. The sixth and seventh columns show the pixel size LpixX and grid width Wproj in nanometers that will result for each combination of number of beamlets per patterned beam, array tilt angle, projection pitch, and K factor. A higher K indicates a faster beamlet deflection scan speed (relative to stage movement), and results in a smaller pixel in the X-direction. If the beamlet control switching signal is supplied to the beamlet blanker array at a constant data rate, the pixel will become bigger in the Y-direction with a relative increase in beamlet deflection scan speed, so that the pixel shape will reduce in the X-direction and enlarge in the Y-direction, changing from approximating a square into a rectangle. The pixels can assume an arbitrary size in the Y-direction by varying the beamlet deflection scan speed, but can only assume particular sizes in the X-direction due to the interleaving of the beamlets as discussed above. In a preferred embodiment of the lithography machine, the machine is simplified by omitting facilities for adjusting individual beamlets. The charged particle dose imparted to the wafer can be lowered by increasing the beamlet deflection scan speed. This lowers the dose even though the beamlet current is unchanged. As a consequence, the pixels will become elongated in the Y-direction. Similarly, the dose can be increased by reducing the beamlet deflection scan speed. FIGS. 17A-17C schematically show yet another embodiment of a method of exposing a target by means of a plurality of beamlets. This method qualifies as a writing strategy of the perpendicular type. In this method, the target is exposed by a group of beamlets in an array by relative movement in the X-direction, while deflecting the group of beamlets in the Y′-direction, such that each beamlet exposes the target along a deflection scan length δY-scan in the Y′-direction. The projection pitch Pproj,Y in the Y′-direction between beamlets is equal to or smaller than the deflection scan length δY-scan of each beamlet in the group of beamlets. In FIG. 17A the first scan line with deflection scan length δY-scan is schematically depicted. In FIG. 17B, exposure of the target has progressed so that each beamlet has exposed six scan lines. In FIG. 17C, exposure has progressed further such that full coverage of an area with length Istripe is obtained beginning from position X1. As can be readily seen, all of the beamlets will be fully utilized to expose the target within the length Istripe. For exposure of a certain length in the Y′-direction, for example L1 in FIG. 13C and Istripe in FIG. 17C, the deflection length per beamlet is shorter for the method of FIG. 17C than for the method of FIG. 12A, 12B or 13A-13C. Furthermore, the beamlet deflection length δY-scan is preferably controlled to avoid stitching errors and the like. The writing strategy of FIG. 17C avoids the “end areas” of partial coverage L2 and L3 shown in FIG. 13C, and thus increases efficiency of the multi-beamlet system. However, the writing strategy of FIG. 17C does not achieve the same beamlet averaging effect. As mentioned earlier, pattern data is generally rasterized with respect to a virtual grid. In order to expose the desired pattern on the target, each beamlet needs to be switched on and off or otherwise modulated in such a way that the sequence of scan lines exposed by the plurality of beamlets form the desired pattern on the target. With respect to the method described with reference to FIGS. 17A-17C, a pattern to be exposed by a single projection lens system of a multi-beam system, e.g. the charged particle multi-beam systems schematically depicted in FIGS. 4 and 5, is divided into blocks with a length of δY-scan. Within each block, the data may be divided into scan lines. Each block is then assigned to beamlets. FIG. 18 is a schematic representation illustrating the effect of a specific orientation of a projection lens array and a beamlet array with respect to each other, i.e. the orientation of plurality of groups of beamlets, and the orientation of the beamlets within each individual group of beamlets. The configuration shown, in fact represents part of an aperture array element of a system according to the invention provided with groups of four apertures, realizing a corresponding number of projections through a single lens system. In the set-up as represented, each group of apertures comprises four beamlets, and the smallest distance between groups, i.e. between apertures of groups is larger than the pitch of the apertures within a group. In this set-up 4×4 groups of apertures have been used, each group for passing beamlets through towards 4×4 projection lenses. The orientation of the beamlets and the projection lenses is such that, during relative movement of the projection lens array and plurality of beamlets with respect to the target in the mechanical scan direction, the entire surface area of the target can be exposed by deflection of the beamlets in a deflection scan direction, generally substantially perpendicular to the mechanical scan direction. In FIG. 18, the orientation of the projection lens array with respect to the mechanical direction of movement is different from the orientation of the beamlet array. The X-direction indicated here is indicative for the movement of a target carrier, while the Y-direction, perpendicular thereto, represents an effective direction of scanning of beamlets over a target as caused by deflection action of deflector 9. FIG. 19 schematically illustrates a scheme to determine the placement of beamlets in a two-dimensional array for use in the method of exposing a target of FIG. 13. In this method, the concept of a diagonal line is used to expose the surface area of the target with an exposure pattern. With this concept, the relative movement in the first direction (e.g. the direction of mechanical movement) and the movement of the plurality of beamlets in the second direction (e.g. the deflection scan direction) are such that the beamlets in the array scan adjacent scan lines in a diagonal pattern. In theory, the group of beamlets could be arranged in an array formed as a diagonal line to perform this writing strategy. In practice, the size of the projections lens systems may be much larger than the desired distance between adjacent scan lines (i.e. the grid cell size) so that arrangement of the beamlets along a diagonal line is not possible. In FIG. 19, it is schematically shown that such a writing strategy will not only work for a diagonal line, but can also be applied when using a suitable two-dimensional array of beamlets. Starting from a diagonal line (left diagram in FIG. 19) of 4 beamlets, rearrangement of the beamlets along the mechanical scan direction (shown in the middle diagram in FIG. 19) leads to a suitable 2×2 array of beamlets (right diagram in FIG. 19). The diagonal line is aligned such that the beamlets are at a projection pitch Pproj,Y in the Y′-direction, and are separated by one scan line interval in the X-direction (one scan line interval being the distance between adjacent scan lines exposed on the target. The 2×2 array of beamlets is then constructed by rearranging the beamlets. Each beamlet in the array is positioned at the same projection pitch Pproj,Y in the Y′-direction as the beamlets in the diagonal line. However, in a direction parallel to the X-direction, each beamlet is displaced from the diagonal line by a whole number of beamlet intervals (one beamlet interval being the distance between subsequent scan lines exposed by the same beamlet). The placement of the beamlets in the array in the X-direction is selected so that the in resulting array, each beamlet is approximately equidistant from at least two other beamlets. The resulting beamlet array is preferably as close as possible to a square, hexagonal, or triangular array. FIGS. 20A-20D schematically illustrate the method of exposing a target by means of a plurality of beamlets using a two-dimensional 2×2 array constructed as shown in FIG. 19. In FIG. 20A, the result of exposure of a portion of a first scan line by each beamlet in a 2×2 array of beamlets is schematically shown. In FIG. 20B, exposure of the target has progressed so that each beamlet has exposed a portion of a second scan line, and in FIG. 20C nine scan lines have been exposed by each beamlet. In FIG. 20D, exposure has progressed further such that an area of full coverage of the target is obtained, shown on the right-hand side of the drawing. From FIGS. 20A-20D it can be seen that the diagonal line writing strategy is applicable. As can be seen in FIG. 20D, in the area of full coverage, the beamlets in the array scan adjacent scan lines in a diagonal pattern. Furthermore, the following may be noticed with respect to this writing strategy. It can be seen that at least some of the scan lines are scanned by more than one beamlet of the array, so that in the example shown, some of the scan lines are scanned over a first portion of the scan line distance by a first beamlet of the array and over a second portion of the scan line distance by a second beamlet of the array. In the example shown, the beamlets are arranged in an array of 2 rows and 2 columns, and the relative movement in the X-direction and movement of the beamlets in the Y′-direction is such that each beamlet in the array scans every second scan line. Furthermore, a portion of the scan lines is scanned by two beamlets, as can be seen on the right-hand side of FIG. 13D where all four beamlets have exposed the target to provide full coverage of the target. If the array of beamlets comprises R rows and C columns of beamlets, the relation between the movements in the X- and Y′-directions is then such that each beamlet in the array exposes every Rth scan line, and that a portion of the scan lines (i.e. in the area where total coverage of the target is achieved) are scanned by C beamlets. Furthermore, if all beamlets of the array are scanned on M adjacent scan lines, each beamlet of the array is scanned on each Mth scan line. This can be seen in the example in FIG. 20D, where all four beamlets are scanned on two adjacent scan lines in the area where total coverage of the target is achieved, and each beamlet of the array is scanned on every second scan line. The writing strategy of FIG. 20D has several advantages. When compared to the writing strategy of FIG. 13C, the “end areas” of partial coverage have been reduced. The length of these end areas is equal to the projection pitch between two beamlets in the Y′-direction Pproj,Y, instead of the projection pitch of the entire array of beamlet as in FIG. 13C. In addition the writing strategy of FIG. 20D achieves averaging of beamlets. In the example shown in FIG. 20D, there is a 2× averaging in the area of full coverage (i.e. two beamlets are used to alternately expose adjacent scan lines). The same writing strategy can be used with a different size array of beamlets to achieve 3× averaging, 4× averaging or more. In embodiments of aforementioned methods, i.e. the methods described with reference to FIGS. 12B and 13A-13C, Pproj,X may be equal to or smaller than a beamlet spot size as projected on the target. Alternatively, the beamlet spot size as projected on the target may be larger than the distance between adjacent scan lines exposed by the plurality of beamlets. The array forming the plurality of beamlets may comprise a number of rows and a number of columns, at least one of the rows and columns being positioned at an angle unequal to 90° with respect to the mechanical scan direction and the deflection scan direction. Alternatively, or additionally, the array forming the plurality of beamlets may be positioned at an angle with respect to the first direction such that the projection pitch between each pair of beamlets in the array in a direction parallel to the first direction is equal. The number of rows may be equal to the number of columns. In embodiments of the invention, the beamlets are charged particle beamlets, preferably electron beamlets. A control unit, e.g. control unit 18, may be arranged to coordinate the relative movement between the target and the plurality of beamlets to enable execution embodiments of aforementioned methods. It should be understood that the control unit, e.g. control unit 18, may be implemented in the form of a computer assembly comprising a processor, and a memory connected to the processor. The memory may comprise memory elements, e.g. a hard disk, Random Access Memory (RAM), Read Only Memory (ROM) Electrically Erasable Programmable Read Only Memory (EEPROM), etc. The processor may be connected to a reading unit, which is arranged to read data from a computer readable medium. The computer readable medium may be arranged to performing, when executed by the processor, embodiments of aforementioned methods of exposing a target by means of a plurality of beamlets in a charged particle multi-beamlet system. Additional aspects of the present invention relate to writing strategies of a perpendicular type. In an embodiment, a method of exposing a target by means of a plurality of beamlets comprises: providing a plurality of beamlets, the beamlets being arranged in an array; converging groups of beamlets towards a common point of convergence for each group providing a target to be exposed; creating relative movement in a first direction between the plurality of beamlets and the target; deflecting each group of beamlets in a second direction substantially perpendicular to the first direction, such that each beamlet within a group of beamlets exposes the target along a deflection scan length in the second direction; wherein a projection pitch Pproj,Y in the second direction between beamlets within a group of beamlets in the array is equal or smaller than the deflection scan length of each beamlet within the group of beamlets. The method may include one or more of the following further features. The projection pitch Pproj,Y may be equal to or smaller than a beamlet spot size as projected on the target. The array of beamlets may comprise a number of rows and a number of columns, at least one of the rows and columns being positioned at an angle unequal to 90° with respect to the first direction and the second direction. A plurality of arrays of beamlets may be provided. The first direction may be substantially perpendicular to the second direction. The array of beamlets form a single patterned beamlet, the single patterned beamlet being patterned in accordance with a rasterized grid with grid cells having a predetermined dimension. In a further aspect, the present invention may encompass a computer readable medium for performing, when executed by a processor, an embodiment of the method of exposing a target by means of a plurality of beamlets in a charged particle multi-beamlet system of the perpendicular type as described above. The invention also encompasses a charged particle multi-beamlet system for patterning a target using a plurality of beamlets, the system comprising: a beamlet pattern generator for providing a radiation pattern formed by a plurality of beamlets, the plurality of beamlets being arranged in groups of beamlets; an array of projection lens systems for projecting the groups of beamlets on to the surface of the target, each project lens system corresponding with a group of beamlets; a deflector array for deflecting a group of beamlets in a second direction substantially perpendicular to the first direction, such that each beamlet within a group of beamlets exposes the target along a modulation scan width in the second direction, the deflector array comprising a plurality of deflectors, each deflector arranged to deflect a corresponding group of beamlets; and a substrate support member for supporting the target to be exposed; wherein a projection pitch Pproj,Y in the second direction between beamlets within a group of beamlets in the array is equal or smaller than the deflection scan width of each beamlet within the group of beamlets; wherein a projection pitch Pproj,Y in the second direction between beamlets within a group of beamlets in the array is equal or smaller than the deflection scan width of each beamlet within the group of beamlets. The charged particle multi-beamlet system may include one or more of the following further features. The projection pitch Pproj,Y may be equal to or smaller than a beamlet spot size as projected on the target. The beamlet pattern generator may comprise: at least one charged particle source for generating a charged particle beam; an aperture array defining separate beamlets or sub-beams from the generated beam; a beamlet manipulator for converging groups of beamlets towards a common point of convergence for each group; and a beamlet blanker for controllably blanking beamlets in the groups of beamlets. The common point of convergence for each group of beamlets may be a point corresponding to one of the projection lens systems. The array of beamlets may comprise a number of rows and a number of columns, at least one of the rows and columns being positioned at an angle unequal to 90° with respect to the first direction and the second direction. The plurality of beamlets may be arranged in a plurality of arrays. The first direction may be substantially perpendicular to the second direction. The plurality of beamlets may form a single patterned beamlet, the single patterned beamlet being patterned in accordance with a rasterized grid with grid cells having a predetermined dimension. Additional aspects of the present invention relate to writing strategies of a diagonal line type. In an embodiment, a method of exposing a target by means of a plurality of beamlets comprises: providing a plurality of beamlets, the beamlets being arranged in an array of N beamlets; providing a target to be exposed; creating relative movement in a first direction between the plurality of beamlets and the target; moving the plurality of beamlets in a second direction such that the beamlets expose a plurality of scan lines on the target; wherein the relative movement in the first direction and the movement of the plurality of beamlets in the second direction are such that the beamlets in the array expose adjacent scan lines in a diagonal pattern. The method may include one or more of the following further features. Each scan line may be scanned in the second direction over a scan line distance, and at least some of the scan lines are scanned by more than one beamlet of the array. At least some of the scan lines may be scanned over a first portion of the scan line distance by a first beamlet of the array and over a second portion of the scan line distance by a second beamlet of the array. All N beamlets of the array may be scanned on M adjacent scan lines, each beamlet of the array being scanned on each Mth scan line. In another embodiment of a writing strategy of a diagonal line type, a method of exposing a target by means of a plurality of beamlets comprises: providing a plurality of beamlets, the beamlets being arranged in an array of R rows and C columns of beamlets; providing a target to be exposed; creating relative movement in a first direction between the plurality of beamlets and the target; moving the plurality of beamlets in a second direction such that the beamlets expose a plurality of scan lines on the target; wherein the relative movement in the first direction and the movement of the plurality of beamlets in the second direction are such that each beamlet in the array scans every Rth scan line, and a portion of the scan lines are scanned by C beamlets. The method may include one or more of the following further features. The plurality of beamlets may be moved in the second direction over a distance of at least approximately R times the projection pitch Pproj,Y in the second direction between beamlets in the array. Each scan line may be scanned in the second direction over a scan line distance, and each beamlet in the array scans a portion of the scan line approximately equal to 1/C of the scan line distance. Each scan line may be scanned by C beamlets of the array of beamlets. In yet another embodiment of a writing strategy of a diagonal line type, a method of exposing a target by means of a plurality of beamlets comprises: providing a plurality of beamlets, the beamlets being arranged in an array; providing a target to be exposed; creating relative movement in a first direction between the plurality of beamlets and the target; and moving the plurality of beamlets in a second direction such that the beamlets expose a plurality of scan lines on the target, adjacent scan lines being separated by a scan line interval and subsequent scan lines exposed by the same beamlet being separated by a beamlet interval; wherein each beamlet in the array is positioned a whole number of beamlet intervals in a direction parallel to the first direction from a diagonal, and at a projection pitch Pproj,Y in the second direction, so that each beamlet is approximately equidistant from at least two other beamlets, the diagonal being aligned with points separated by one scan line interval in the first direction and by the projection pitch Pproj,Y in the second direction. The embodiments of a writing strategy of the diagonal line type as described above may include one or more of the following features. A projection pitch Pproj,X in the first direction between beamlets of the array may be equal to or smaller than a beamlet spot size as projected on the target. The array of beamlets may comprise a number of rows and a number of columns, at least one of the rows and columns being positioned at an angle unequal to 90° with respect to the first direction and the second direction. A plurality of arrays of beamlets may be provided. The first direction may be substantially perpendicular to the second direction. The array of beamlets may form a single patterned beamlet, the single patterned beamlet being patterned in accordance with a rasterized grid with grid cells having a predetermined dimension. In a further aspect, the present invention may encompass a computer readable medium for performing, when executed by a processor, an embodiment of the method of exposing a target by means of a plurality of beamlets in a charged particle multi-beamlet system of the diagonal line type as described above. The invention also encompasses a charged particle multi-beamlet system for patterning a target using a plurality of beamlets. The charged particle multi-beamlet system comprises: a beamlet pattern generator for providing a radiation pattern formed by a plurality of beamlets, the plurality of beamlets being arranged in an array of N beamlets; an array of projection lens systems for projecting the plurality of beamlets on to the surface of the target, each project lens system corresponding with a group of beamlets; a deflector array for deflecting a group of beamlets in a second direction substantially perpendicular to the first direction, such that each beamlet within a group of beamlets exposes the target along a modulation scan width in the second direction, the deflector array comprising a plurality of deflectors, each deflector arranged to deflect a corresponding group of beamlets; a substrate support member for supporting the target to be exposed; and a control unit arranged to coordinate relative movement between the substrate support member and the plurality of beamlets in a first direction and movement of the group of beamlets in the second direction such that the beamlets in the array scan adjacent scan lines in a diagonal pattern. The charged particle multi-beamlet system may include one or more of the following further features. The control unit may be further arranged to control scanning in the second direction over a scan line distance, and control that at least some of the scan lines are scanned by more than one beamlet of the array. The control unit may be further arranged to control that at least some of the scan lines are scanned over a first portion of the scanline distance by a first beamlet of the array and over a second portion of the scanline distance by a second beamlet of the array. All N beamlets of the array may be scanned on M adjacent scan lines, each beamlet of the array being scanned on each Mth scan line. In another embodiment, a charged particle multi-beamlet system for patterning a target using a plurality of beamlets comprises: a beamlet pattern generator for providing a radiation pattern formed by a plurality of beamlets, the plurality of beamlets being arranged in groups of beamlets, the group of beamlets being arranged in an array of R rows and C columns of beamlets; an array of projection lens systems for projecting the groups of beamlets on to the surface of the target, each project lens system corresponding with a group of beamlets; a deflector array for deflecting a group of beamlets in a second direction substantially perpendicular to the first direction, such that each beamlet within a group of beamlets exposes the target along a modulation scan width in the second direction, the deflector array comprising a plurality of deflectors, each deflector arranged to deflect a corresponding group of beamlets; a substrate support member for supporting the target to be exposed; and a control unit arranged to coordinate relative movement between the substrate support member and the plurality of beamlets in a first direction and movement of the group of beamlets in the second direction such that each beamlet in the array scans every Rth scan line, and a portion of the scan lines are scanned by C beamlets. The charged particle multi-beamlet system may include one or more of the following further features. The control unit may be further arranged to move the plurality of beamlets in the second direction over a distance of at least approximately R times the projection pitch Pproj,Y in the second direction between beamlets in the array. The control unit may be further arranged to coordinate the relative movement between the substrate support member and the plurality of beamlets in the first direction and the movement of the group of beamlets in the second direction such that each scan line is scanned in the second direction over a scan line distance, and each beamlet in the array scans a portion of the scan line approximately equal to 1/C of the scan line distance. Each scan line may be scanned by C beamlets of the array of beamlets. In yet another embodiment, a charged particle multi-beamlet system for patterning a target using a plurality of beamlets comprises: a charged particle source for generating a charged particle beam; a beamlet aperture array for defining groups of beamlets from the generated beam; a deflector array for deflecting a group of beamlets in a second direction, the deflector array comprising a plurality of deflectors, each deflector arranged to deflect a corresponding group of beamlets; an array of projection lens systems for projecting the groups of beamlets on to the surface of the target, each project lens system corresponding with a group of beamlets; a substrate support for supporting the target to be exposed; and a control unit arranged to coordinate relative movement between the substrate support and the plurality of beamlets in a first direction and movement of the group of beamlets in the second direction such that the beamlets expose a plurality of scan lines on the target, adjacent scan lines being separated by a scan line interval and subsequent scan lines exposed by the same beamlet being separated by a beamlet interval; wherein each group of beamlets projected onto the target is arranged in an array, each beamlet in the array being positioned a whole number of beamlet intervals in a direction parallel to the first direction from a diagonal, and at a projection pitch Pproj,Y in the second direction, so that each beamlet is approximately equidistant from at least two other beamlets, the diagonal being aligned with points separated by one scan line interval in the first direction and by the projection pitch Pproj,Y in the second direction. The embodiments of a charged particle multi-beamlet system as described above may include one or more of the following features. The projection pitch Pproj,X may be equal to or smaller than a beamlet spot size as projected on the target. The beamlet pattern generator may comprise: at least one charged particle source for generating a charged particle beam; an aperture array defining separate beamlets or sub-beams from the generated beam; a beamlet manipulator for converging groups of beamlets towards a common point of convergence for each group; and a beamlet blanker for controllably blanking beamlets in the groups of beamlets. The common point of convergence for each group of beamlets may be a point corresponding to one of the projection lens systems. The array of beamlets may comprise a number of rows and a number of columns, at least one of the rows and columns being positioned at an angle unequal to 90° with respect to the first direction and the second direction. the plurality of beamlets are arranged in a plurality of arrays. The first direction may be substantially perpendicular to the second direction. The plurality of beamlets may form a single patterned beamlet, the single patterned beamlet being patterned in accordance with a rasterized grid with grid cells having a predetermined dimension. The invention has been described by reference to certain embodiments discussed above. It will be recognized that these embodiments are susceptible to various modifications and alternative forms well known to those of skill in the art without departing from the spirit and scope of the invention. Accordingly, although specific embodiments have been described, these are examples only and are not limiting upon the scope of the invention, which is defined in the accompanying claims.
	description	This application claims priority under 35 U.S.C. § 119 to Korean Patent Application No. 10-2014-0103110, filed on Aug. 11, 2014, in the Korean Intellectual Property Office, the disclosure of which is incorporated herein by reference in its entirety. The following disclosure relates to a reactor and an operating method for the reactor, and more particularly, to a reactor which may passively cool excessively generated heat without an operation of an operator at the time of abnormality of the reactor, completely passively perform the cooling operation for safety procedures by a structure of the reactor and a change in environmental conditions such as a pressure, etc., without a separate control command, and have a simpler structure than the existing reactor safety system, and an operating method for the reactor. Nuclear power generation is made by a method for turning a turbine using energy generated at the time of nuclear fission to generate electric energy. FIG. 1 schematically illustrates a principle of typical nuclear power generation. An enormous amount of heat energy is generated by the nuclear fission of nuclear fuel within a pressure vessel (or referred to as a reactor vessel). The heat energy is transferred to a coolant within the pressure vessel and the coolant is circulated to be released from the pressure vessel and to be reintroduced into the pressure vessel via a heat exchanger, as illustrated by a thick arrow in FIG. 1. The heat energy of the coolant is transferred to a steam generator while passing through the heat exchanger and water within the steam generator is converted into high-temperature and high-pressure steam due to the heat energy. The so generated high-temperature and high-pressure steam is supplied to the turbine as illustrated by a thin arrow in FIG. 1, the turbine rotates by power of steam, a generator connected to the turbine rotates together, thereby generating power. The steam from which the energy is lost by the rotation of the turbine again goes through the phase change to be converted into water. As illustrated by the thin arrow in FIG. 1, the water is reintroduced into the steam generator and is thus circulated. FIG. 1 illustrates only systems which are a subject of the nuclear power generator. Actually, the reactor is essentially provided with the safety system. As described above, very high heat is generated when the reactor is operated. The high-heat environment includes a very high risk to cause disaster when the reactor is damaged. Therefore, when the reactor is damaged, the safety system should be essentially provided to quickly cool the reactor. Therefore, the related art includes various types of safety systems which may safely cool the reactor at the time of the accident of the reactor. Describing in more detail, the safety system which is applied to the existing reactor may include components (ex. passive residual heat removal (PRHR) system, etc.) which circulate the coolant accommodated in the reactor vessel to the outside, components (ex. core make-up tank (CMT), safety injection pump (SI pump), etc.) which supply the coolant separately accommodated in the outside to the vessel), etc. The example is illustrated in FIG. 2. Further, all of the safety systems of the foregoing example use the coolant (water) to perform the cooling. To more increase cooling efficiency, the configuration of the safety systems which simultaneously use air and water is also disclosed. As an example of the safety system technology of diversifying a heat sink using air and water, there are Japanese Patent Laid-Open Publication No. 2010-217091 (“Containment Vessel Passive Cooling System And Liquid-Metal-Cooled-Reactor”), Korean Patent Laid-Open Publication No. 2009-0021722 (“Air/Water Hybrid Passive Reactor Cavity Cooling Apparatus And Method For Core Decay Heat Removal Of High Temperature Gas-Cooled Reactor”), etc. However, research and development for more improving efficiency of a reactor driving system have been continuously conducted and therefore only the typical safety system may not obtain the sufficient cooling efficiency. Further, if the safety system is configured to be operated by receiving a separate control command even at the moment of urgent accident like the damage of the reactor, there are many risks that the safety system malfunctions due to the damage of the control system and thus the safety system is not properly operated, an operator of the reactor does not issue the control command in time, etc. In addition, as described in the foregoing related arts, since the structure of the reactor safety system is very complicated and multiplexed, design factors to be considered to construct facilities are too many and thus it is very difficult to design and actually construct the safety system and even after the facility construction is completed, ones to be tested or monitored are too many for safe operation and thus it is difficult to operate and control the safety system. Therefore, a need exists for the reactor safety system which may more improve the cooling efficiency than that of the existing safety system, may be completely passively operated without performing the separate control operation of the operator, and may have a simpler structure than the existing reactor safety system. 1. Japanese Patent Laid-Open Publication No. 2010-217091 (“Containment Vessel Passive Cooling System And Liquid-Metal-Cooled-Reactor”) 2. Korean Patent Laid-Open Publication No. 2009-0021722 (“Air/Water Hybrid Passive Reactor Cavity Cooling Apparatus And Method For Core Decay Heat Removal Of High Temperature Gas-Cooled Reactor”) An embodiment of the present invention is directed to providing a reactor which may passively cool excessively generated heat without an operation of an operator at the time of abnormality of the reactor, completely passively perform the cooling operation for safety procedures by a structure of the reactor and a change in environmental conditions such as a pressure, etc., without a separate control command, and have a simpler structure than the existing reactor safety system, and an operating method for the reactor. In one general aspect, a reactor includes: a reactor driving system configured to include a reactor vessel accommodating a reactor core and a steam generator to which a steam pipe and a water supply pipe are connected; and a reactor safety system configured by being divided into an energy release space (ESR) accommodating the reactor driving system, an energy absorbing space (EAS) communicating with the energy release space through a passage formed thereover and accommodating a coolant, and an energy transfer space (ETS) formed to be isolated from the energy release space and the energy absorbing space and having a heat exchange device provided therein to transfer heat released from the reactor driving system to the coolant, the heat exchange device being connected to the energy release space and the energy absorbing space, respectively, wherein the coolant within the reactor safety system is selectively distributed in response to thermal-hydraulic conditions changed depending on a change in pressure within the reactor driving system and whether the coolant is leaked to cool the reactor driving system. In another general aspect, a reactor includes: a reactor driving system configured to include a reactor vessel 112 accommodating a reactor core 111 and a steam generator 113 to which a steam pipe 114 and a water supply pipe 115 are connected; and a reactor safety system configured to include a releasing isolation vessel 120 accommodating gas and the reactor driving system, an absorbing isolation vessel 130 communicating with the releasing isolation vessel 120 through a passage 123 formed over the releasing isolation vessel 120 and accommodating a coolant, a transferring isolation vessel 140 provided over the absorbing isolation vessel 130 to accommodate the gas and the coolant, a releasing heat exchanger 121 and an absorbing heat exchanger 131 adjacently disposed to each other within the transferring isolation vessel 140 to exchange heat with each other, a coolant spray pipe 134 spraying the coolant to the releasing heat exchanger 121 and the absorbing heat exchanger 131 adjacently disposed to each other, a condensing heat exchanger 133 disposed outside the isolation vessels, a releasing heat exchange channel 122 connected to the reactor vessel 112 and the releasing heat exchanger 121 to circularly distribute the coolant, and an absorbing heat exchange channel 132 connected to the absorbing heat exchanger 131 and the condensing heat exchanger 133 to circularly distribute the coolant, wherein the coolant within the reactor safety system is selectively distributed in response to thermal-hydraulic conditions changed depending on a change in pressure within the reactor driving system and whether the coolant is leaked to cool the reactor driving system. In the reactor safety system, the coolant sprayed from the coolant spray pipe 134 may absorb heat from the coolant distributed within the releasing heat exchanger 121 to be evaporated, the coolant distributed within the absorbing heat exchanger 131 may absorb the heat of the steam generated by the evaporation to be condensed and may be formed to be accommodated into the transferring isolation vessel 140, and the heat transfer may be made by the two-phase heat transfer mechanism for transferring heat from the coolant within the releasing heat exchanger 121 to the coolant within the absorbing heat exchanger 131 by evaporating and the condensing the steam sprayed by the coolant spray pipe 134. The coolant spray pipe 134 may have one end communicating with the absorbing isolation vessel 130 to be supplied with the coolant and the other end provided with the nozzle to spray the supplied coolant to the releasing heat exchanger 121 and the absorbing heat exchanger 131 and is provided with a coolant spray valve 134a. The reactor safety system may further include: a coolant injection pipe 135 having one end communicating with the transferring isolation vessel 140 and the other end communicating with the releasing isolation vessel 120 to inject the coolant accommodated in the transferring isolation vessel 140 into the releasing isolation vessel 120; and a coolant injection valve 135a provided on the coolant injection pipe 135. The transferring isolation vessel 140 may have an accommodating barrier 141 enclosing an area in which the releasing heat exchanger 121 and the absorbing heat exchanger 131 are disposed to accommodate the cold water within the transferring isolation vessel 140. The passage 123 may be formed in a double barrier form in which a releasing isolation vessel barrier 123a is formed at the releasing isolation vessel 120 and has an upper portion opened and an absorbing isolation vessel barrier 123b is formed at the absorbing isolation vessel 130 and has a lower portion opened. The reactor safety system may further include: a releasing isolation vessel communicating pipe 124 formed to communicate between the upper portion of the releasing isolation vessel 120 and the upper portion of the transferring isolation vessel 140; and a releasing isolation vessel communicating valve 124a provided on the releasing isolation vessel communicating pipe 124. The reactor safety system may further include: a releasing isolation vessel pressure reducing pipe 125 formed to communicate between a lower portion of the releasing isolation vessel 120 and a lower portion of the absorbing isolation vessel 130; and a releasing isolation vessel pressure reducing valve 125a provided on the releasing isolation vessel pressure reducing pipe 125. The reactor safety system may further include: a steam bypass pipe 126 having one end communicating with the steam pipe 114 and the other end communicating with the releasing heat exchange channel 122 to distribute the steam generated by the evaporation of the coolant within the steam generator 113 to the releasing heat exchange channel 122; and a steam bypass valve 126a provided on the steam bypass pipe 126. The reactor safety system may further include: a steam release pipe having one end communicating with the steam pipe 114 and the other end communicating with the space within the releasing isolation vessel 120 to release the steam generated by the evaporation of the coolant within the reactor vessel 112 to the space within the releasing isolation vessel 120; and a steam release valve 127a provided on the steam release pipe. The reactor safety system may further include: a coolant supplement pipe 136 having one end communicating with the releasing heat exchange channel 122 and the other end communicating with a space under the surface of the coolant within the absorbing isolation vessel 130 or the transferring isolation vessel 140 to supplement the coolant to the releasing heat exchange channel 122 and a coolant supplement valve 136a provided on the coolant supplement pipe 136. The reactor safety system may further include: an auxiliary waste heat removing part 150 including an auxiliary absorbing heat exchanger 151 disposed within the absorbing isolation vessel 130, an auxiliary condensing heat exchanger 153 disposed outside the isolation vessels, and an auxiliary absorbing heat exchange channel 152 connected to the auxiliary absorbing heat exchanger 151 and the auxiliary condensing heat exchanger 153 to circularly distribute the coolant. In another general aspect, an operating method for a reactor operates the reactor configured as described above and the operating method includes: isolating which includes stopping a reactor driving system and closing a steam pipe isolation valve 114a included in a steam pipe 114 and a water supply pipe isolation valve 115a included in a water supply pipe 115; circulating the coolant into a releasing heat exchanger which includes introducing steam generated by evaporation of a coolant within a reactor vessel 112 into the releasing heat exchanger 121 through a releasing heat exchange channel 122; and generating the cooler by condensing steam introduced into the releasing heat exchanger 121 by passing the stream through the releasing heat exchanger 121 and circulating the generated coolant by reintroducing the generated coolant into the reactor vessel 112 through the releasing heat exchange channel 122; transferring a two-phase heat which includes spraying the coolant supplied by a coolant spray pipe 134 to an outer surface of the releasing heat exchanger 121 to contact the outer surface of the releasing heat exchanger 121; evaporating the coolant contacting the outer surface of the releasing heat exchanger 121 by allowing the coolant to absorb heat from the steam within the releasing heat exchanger 121 and condensing the steam within the releasing heat exchanger 121 to generate the coolant; allowing the steam generated by the evaporation at the outer surface of the releasing heat exchanger 121 to contact the outer surface of the absorbing heat exchanger 131; allowing the steam contacting the outer surface of the absorbing heat exchanger 131 to release heat to the coolant within the absorbing heat exchanger 131 to be condensed by the coolant and evaporating the coolant within the absorbing heat exchanger 131 to generate the steam; and circulating the absorbing heat exchanger which includes introducing the steam within the absorbing heat exchanger 11 into the condensing heat exchanger 133 through the absorbing heat exchange channel 131, condensing the steam introduced into the condensing heat exchanger 133 by passing the steam through the condensing heat exchanger 133 to generate the coolant, and circulating the generated coolant by reintroducing the generated coolant into the absorbing heat exchanger 131 through the absorbing heat exchange channel 132. The coolant spray pipe 134 may have one end communicating with the absorbing isolation vessel 130 to be supplied with the coolant and the other end provided with the nozzle to spray the supplied coolant to the releasing heat exchanger 121 and the absorbing heat exchanger 131 and is provided with a coolant spray valve 134a, and the reactor safety system may further include: a coolant injection pipe 135 having one end communicating with the transferring isolation vessel 140 and the other end communicating with the releasing isolation vessel 120 to inject the coolant accommodated in the transferring isolation vessel 140 into the releasing isolation vessel 120; and a coolant injection valve 135a provided on the coolant injection pipe 135, and the operating method for a reactor further include: pressurizing the releasing isolation vessel 120 and spraying the coolant which includes leaking the steam generated by the evaporation of the coolant within the reactor vessel 112 to fill the space within the releasing isolation vessel 120 so as to pressurize the releasing isolation vessel 120; transferring the pressure to the coolant within the absorbing isolation vessel 130 through the passage 1230 to introduce the coolant into the coolant spray pipe 134; and opening the coolant spray valve 134a by the pressure to spray the coolant to the releasing heat exchanger 121 and the absorbing heat exchanger 131 through the coolant spray pipe 134; and directly cooling the coolant which includes accommodating the coolant generated by condensing steam by performing the two-phase heat transfer into the transferring isolation vessel 140; introducing the coolant accommodated in the transferring isolation vessel 140 into the coolant injection pipe 135; and opening the coolant injection valve 135a by the pressure to inject the coolant into the releasing isolation vessel 120 through the coolant injection pipe 135 so that the coolant directly contacts the reactor vessel 112 to perform cooling. The reactor safety system may further include: a releasing isolation vessel communicating pipe 124 formed to communicate between the upper portion of the releasing isolation vessel 120 and the upper portion of the transferring isolation vessel 140; and a releasing isolation vessel communicating valve 124a provided on the releasing isolation vessel communicating pipe 124, and the operating method for a reactor may further include: leaking the steam generated by the evaporation of the coolant within the reactor vessel 112 to fill the space within the releasing isolation vessel 120 so as to pressurize the releasing isolation vessel 120; and opening the releasing isolation vessel communicating valve 124a by the pressure to releasing the steam within the releasing isolation vessel 120 to the transferring isolation vessel 140 through the releasing isolation vessel communicating pipe 124 so as to fill and pressurize the space within the transferring isolation vessel 140, such that auxiliary pressurization for opening the coolant injection valve 135a in the direct cooling of the coolant is additionally performed. The reactor safety system may further include: a releasing isolation vessel pressure reducing pipe 125 formed to communicate between a lower portion of the releasing isolation vessel 120 and a lower portion of the absorbing isolation vessel 130; and a releasing isolation vessel pressure reducing valve 125a provided on the releasing isolation vessel pressure reducing pipe 125, and the operating method for a reactor may further include leaking the steam generated by the evaporation of the coolant within the reactor vessel 112 to fill the space within the releasing isolation vessel 120 so as to pressurize the releasing isolation vessel 120; filling the releasing isolation vessel 120 with the coolant by the direct cooling of the coolant; and opening the releasing isolation vessel pressure reducing valve 125a by the pressure to release the coolant within the releasing isolation vessel 120 to the absorbing isolation vessel 130 through the releasing isolation vessel pressure reducing pipe 125, such that the pressure reduction of the overpressure formed within the releasing isolation vessel 120 is additionally performed. The reactor safety system may further include a steam bypass pipe 126 having one end communicating with the steam pipe 114 and the other end communicating with the releasing heat exchange channel 122 to distribute the steam generated by the evaporation of the coolant within the steam generator 113 to the releasing heat exchange channel 122; and a steam bypass valve 126a provided on the steam bypass pipe 126, and the operating method for a reactor may further include pressurizing the space within the steam generator 113 by the steam generated by the evaporation of the coolant within the steam generator 113; and opening the steam bypass valve 126a by the pressure to introduce the steam within the steam generator 113 into the releasing heat exchange channel 122 through the steam bypass pipe 126, such that the steam auxiliary introduction into the releasing heat exchange channel 122 is additionally performed. The reactor safety system may further include a steam release pipe having one end communicating with the steam pipe 114 and the other end communicating with the space within the releasing isolation vessel 120 to release the steam generated by the evaporation of the coolant within the reactor vessel 112 to the space within the releasing isolation vessel 120 and a steam release valve 127a provided on the steam release pipe, and the operating method for a reactor may further include pressurizing the space within the steam generator 113 by the steam generated by the evaporation of the coolant within the steam generator 113; and opening the steam release valve 127a by the pressure to release the steam within the steam generator 113 to the releasing isolation vessel 120 through the steam release pipe, such that the auxiliary pressurization within the releasing isolation vessel 120 is be additionally performed. The reactor safety system may further include a coolant supplement pipe 136 having one end communicating with the releasing heat exchange channel 122 and the other end communicating with a space under the surface of the coolant within the absorbing isolation vessel 130 or the transferring isolation vessel 140 to supplement the coolant to the releasing heat exchange channel 122 and a coolant supplement valve 136a provided on the coolant supplement pipe 136, and the operating method for a reactor may further include opening the coolant supplement valve 136a by the pressure to introduce the coolant into the releasing heat exchange channel 122 through the coolant supplement pipe 136, such that the auxiliary introduction of the coolant into the releasing heat exchange channel 122 may be additionally performed. The reactor safety system may further include an auxiliary waste heat removing part 150 which includes an auxiliary absorbing heat exchanger 151 disposed within the absorbing isolation vessel 130, an auxiliary condensing heat exchanger 153 disposed outside the isolation vessels, and an auxiliary absorbing heat exchange channel 152 connected to the auxiliary absorbing heat exchanger 151 and the auxiliary condensing heat exchanger 153 to circularly distribute the coolant, and the operating method for a reactor may further include removing auxiliary waste heat which includes absorbing, by a coolant within the auxiliary absorbing heat exchanger 151, heat from the coolant within the absorbing isolation vessel 130, introducing the coolant within the auxiliary absorbing heat exchanger 151 into the auxiliary condensing heat exchanger 153 through the auxiliary absorbing heat exchange channel 152, releasing the coolant within the auxiliary condensing heat exchanger 153 to external environment, and reintroducing the coolant within the auxiliary condensing heat exchanger 153 into the auxiliary absorbing heat exchanger 151 through the auxiliary absorbing heat exchange channel 152 to be circulated. In another general aspect, an operating method for a reactor operates the reactor configured as described above and the operating method includes: isolating which includes stopping a reactor driving system and closing a steam pipe isolation valve 114a included in a steam pipe 114 and a water supply pipe isolation valve 115a included in a water supply pipe 115; circulating the coolant into a releasing heat exchanger which includes introducing steam generated by evaporation of a coolant within a reactor vessel 112 into the releasing heat exchanger 121 through a releasing heat exchange channel 122; and generating the coolant by condensing steam introduced into the releasing heat exchanger 121 by passing the steam through the releasing heat exchanger 121 and circulating the generated coolant by reintroducing the generated coolant into the reactor vessel 112 through the releasing heat exchange channel 122; performing a two-phase heat transfer which includes spraying the coolant supplied by a coolant spray pipe 134 to an outer surface of the releasing heat exchanger 121 to contact the outer surface of the releasing heat exchanger 121; evaporating the coolant contacting the outer surface of the releasing heat exchanger 121 by allowing the coolant to absorb heat from the steam within the releasing heat exchanger 121 and condensing the steam within the releasing heat exchanger 121 to generate the coolant; allowing the steam generated by the evaporation at the outer surface of the releasing heat exchanger 121 to contact the outer surface of the absorbing heat exchanger 131; allowing the steam contacting the outer surface of the absorbing heat exchanger 131 to release heat to the coolant within the absorbing heat exchanger 131 to be condensed to the coolant and evaporating the coolant within the absorbing heat exchanger 131 to generate the steam; and circulating the absorbing heat exchanger which includes introducing the steam within the absorbing heat exchanger 131 into the condensing heat exchanger 133 through the absorbing heat exchange channel 132, condensing the steam introduced into the condensing heat exchanger 133 by passing the steam through the condensing heat exchanger 133 to generate the coolant, and circulating the generated coolant by reintroducing the generated coolant into the absorbing heat exchanger 131 through the absorbing heat exchange channel 132. The coolant spray pipe 134 may have one end communicating with the absorbing isolation vessel 130 to be supplied with the coolant and the other end provided with the nozzle to spray the supplied coolant to the releasing heat exchanger 121 and the absorbing heat exchanger 131 and is provided with a coolant spray valve 134a, and the reactor safety system may further include: a coolant injection pipe 135 having one end communicating with the transferring isolation vessel 140 and the other end communicating with the releasing isolation vessel 120 to inject the coolant accommodated in the transferring isolation vessel 140 into the releasing isolation vessel 120; and a coolant injection valve 135a provided on the coolant injection pipe 135, and the operating method for a reactor further include: pressurizing the releasing isolation vessel 120 and spraying the coolant which includes leaking the steam generated by the evaporation of the coolant within the reactor vessel 112 to fill the space within the releasing isolation vessel 120 so as to pressurize the releasing isolation vessel 120; transferring the pressure to the coolant within the absorbing isolation vessel 130 through the passage 123 to introduce the coolant into the coolant spray pipe 134; and opening the coolant spray valve 134a by the pressure to spray the coolant to the releasing heat exchanger 121 and the absorbing heat exchanger 131 through the coolant spray pipe 134; and directly cooling the coolant which includes accommodating the coolant generated by condense the steam by the performing of the two-phase heat transfer into the transferring isolation vessel 140; introducing the coolant accommodated in the transferring isolation vessel 140 into the coolant injection pipe 135; and opening the coolant injection valve 135a by the pressure to inject the coolant into the releasing isolation vessel 120 through the coolant injection pipe 135 so that the coolant directly contacts the reactor vessel 112 to perform cooling. The reactor safety system may further include: a releasing isolation vessel communicating pipe 124 formed to communicate between the upper portion of the releasing isolation vessel 120 and the upper portion of the transferring isolation vessel 140; and a releasing isolation vessel communicating valve 124a provided on the releasing isolation vessel communicating pipe 124, and the operating method for a reactor may further include: leaking the steam generated by the evaporation of the coolant within the reactor vessel 112 to fill the space within the releasing isolation vessel 120 so as to pressurize the releasing isolation vessel 120; and opening the releasing isolation vessel communicating valve 124a by the pressure to release the steam within the releasing isolation vessel 120 to the transferring isolation vessel 140 through the releasing isolation vessel communicating pipe 124 so as to fill and pressurize the space within the transferring isolation vessel 140, such that auxiliary pressurization for opening the coolant injection valve 135a in the direct cooling of the coolant is additionally performed. The reactor safety system may further include: a releasing isolation vessel pressure reducing pipe 125 formed to communicate between a lower portion of the releasing isolation vessel 120 and a lower portion of the absorbing isolation vessel 130; and a releasing isolation vessel pressure reducing valve 125a provided on the releasing isolation vessel pressure reducing pipe 125, and the operating method for a reactor may further include leaking the steam generated by the evaporation of the coolant within the reactor vessel 112 to fill the space within the releasing isolation vessel 120 so as to pressurize the releasing isolation vessel 120; filling the releasing isolation vessel 120 with the coolant by the direct cooling of the coolant; and opening the releasing isolation vessel pressure reducing valve 125a by the pressure to discharge the coolant within the releasing isolation vessel 120 to the absorbing isolation vessel 130 through the releasing isolation vessel pressure reducing pipe 125, such that the pressure reduction of the overpressure formed within the releasing isolation vessel 120 is additionally performed. The reactor safety system may further include a steam bypass pipe 126 having one end communicating with the steam pipe 114 and the other end communicating with the releasing heat exchange channel 122 to distribute the steam generated by the evaporation of the coolant within the steam generator 113 to the releasing heat exchange channel 122; and a steam bypass valve 126a provided on the steam bypass pipe 126, and the operating method for a reactor may further include pressurizing the space within the steam generator 113 by the steam generated by the evaporation of the coolant within the steam generator 113; and opening the steam bypass valve 126a by the pressure to introduce the steam within the steam generator 113 into the releasing heat exchange channel 122 through the steam bypass pipe 126, such that the steam auxiliary introduction into the releasing heat exchange channel 122 is additionally performed. The reactor safety system may further include a steam release pipe having one end communicating with the steam pipe 114 and the other end communicating with the space within the releasing isolation vessel 120 to release the steam generated by the evaporation of the coolant within the reactor vessel 112 to the space within the releasing isolation vessel 120 and a steam release valve 127a provided on the steam release pipe, and the operating method for a reactor may further include pressurizing the space within the steam generator 113 by the steam generated by the evaporation of the coolant within the steam generator 113; and opening the steam release valve 127a by the pressure to release the steam within the steam generator 113 to the space within the releasing isolation vessel 120 through the steam release pipe, such that the auxiliary pressurization within the releasing isolation vessel 120 may be additionally performed. The reactor safety system may further include a coolant supplement pipe 136 having one end communicating with the releasing heat exchange channel 122 and the other end communicating with a space under the surface of the coolant within the absorbing isolation vessel 130 or the transferring isolation vessel 140 to supplement the coolant to the releasing heat exchange channel 122 and a coolant supplement valve 136a provided on the coolant supplement pipe 136, and the operating method for a reactor may further include opening the coolant supplement valve 136a by the pressure to introduce the coolant into the releasing heat exchange channel 122 through the coolant supplement pipe 136, such that the auxiliary introduction of the coolant into the releasing heat exchange channel 122 may be additionally performed. The reactor safety system may further include an auxiliary waste heat removing part 150 which includes an auxiliary absorbing heat exchanger 151 disposed within the absorbing isolation vessel 130, an auxiliary condensing heat exchanger 153 disposed outside the isolation vessels, and an auxiliary absorbing heat exchange channel 152 connected to the auxiliary absorbing heat exchanger 151 and the auxiliary condensing heat exchanger 153 to circularly distribute the coolant, and the operating method for a reactor may further include removing auxiliary waste heat which includes absorbing, by a coolant within the auxiliary absorbing heat exchanger 151, heat from the coolant within the absorbing isolation vessel 130, introducing the coolant within the auxiliary absorbing heat exchanger 151 into the auxiliary condensing heat exchanger 153 through the auxiliary absorbing heat exchange channel 152, releasing, the coolant within the auxiliary condensing heat exchanger 153, heat to external environment, and reintroducing the coolant within the auxiliary condensing heat exchanger 153 into the auxiliary absorbing heat exchanger 151 through the auxiliary absorbing heat exchange channel 152 to be circulated. Other features and aspects will be apparent from the following detailed description, the drawings, and the claims. 100: (The inventive) reactor111: Reactor core112: Reactor vessel113: Steam generator114: Steam pipe114a: Steam pipe isolation valve115: Water supply pipe115a: Water supply pipe isolation valve120: Releasing isolation vessel121: Releasing heat exchanger122: Releasing heat exchange channel122a: Reactor safety valve123: Passage123a: Releasing isolation vessel barrier123b: Absorbing isolation vessel barrier124: Releasing isolation vessel communicating pipe124a: Releasing isolation vessel communicating valve125: Releasing isolation vessel pressure reducing pipe125a: Releasing isolation vessel pressure reducing valve126: Steam bypass pipe126a: Steam bypass valve127: Steam release pipe127a: Steam release valve128: Water supply supplement pipe128a: Water supply supplement valve130: Absorbing isolation vessel131: Absorbing heat exchanger132: Absorbing heat exchange channel133: Condensing heat exchanger134: Coolant spray pipe134a: Coolant spray valve135: Coolant injection pipe135a: Coolant injection valve136: Coolant supplement pipe136a: Coolant supplement valve140: Transferring isolation vessel141: Accommodating barrier150: Auxiliary waste heat removing part151: Auxiliary absorbing heat exchanger152: Auxiliary condensing heat exchanger153: Auxiliary absorbing heat exchange channel Hereinafter, a reactor and an operating method for the reactor according to an exemplary embodiment of the present invention having the above-mentioned configuration will be described in detail with reference to the accompanying drawings. FIG. 3 is a conceptual diagram of a structure of a reactor according to an exemplary embodiment of the present invention and FIGS. 4A, 4B, and 4C are diagrams illustrating an operation principle of the conceptual reactor according to the exemplary embodiment of the present invention as illustrated in FIG. 3. The reactor according to the exemplary embodiment of the present invention has a reactor safety system which may quickly and effectively perform cooling when a reactor driving system stops due to accidents, etc. Therefore, the reactor driving system according to the exemplary embodiment of the present invention may be configured in any form. Describing in detail, the reactor driving system may be a reactor driving system having a form which is generally used in a large reactor, that is, a form in which a reactor vessel accommodating a reactor core and a steam generator are separately provided while being spaced apart from each other or a form which is used in a small reactor, that is, a reactor driving system in which the steam generator is integrally accommodated in the reactor vessel. Whatever form the reactor driving system takes (that is, independent of the driving system for the large reactor or the small reactor), the reactor driving system basically includes the reactor vessel accommodating the reactor core and the steam generator to which a steam pipe and a water supply pipe are connected. As illustrated in FIG. 3, the reactor safety system according to the exemplary embodiment of the present invention is largely divided into an energy release space (ERS), an energy absorbing space (EAS), and an energy transfer space (ETS). The energy release space (ERS) accommodates the reactor driving system and the energy absorbing space (EAS) accommodates a coolant. As illustrated, the energy absorbing space (EAS) is configured to communicate with the energy release space (ERS) by a passage which is formed thereover. Further, the energy transfer space (ETS) is configured to be isolated from the energy release space (ERS) and the energy absorbing space (EAS). The energy transfer space (ETS) has a heat exchange device which is connected to the energy release space (ERS) and the energy absorbing space (EAS), respectively, to transfer heat released from the reactor driving system to the coolant. The reactor safety system according to the exemplary embodiment of the present invention is configured to selectively distribute the coolant within the reactor safety system in response to thermal-hydraulic conditions changed depending on a change in pressure within the reactor driving system and whether the coolant is leaked, to thereby cool the reactor driving system. An operation principle of the conceptual configuration of the reactor according to the exemplary embodiment of the present invention will be described in more detail with reference to FIG. 4A, 4B or 4C. FIG. 4A is a diagram illustrating a normal state of the conceptual configuration of the reactor according to the exemplary embodiment of the present invention as illustrated in FIG. 3. When the reactor is normally operated, there is no need to additionally cool the reactor, such that the reactor is maintained in the state of FIG. 4A. That is, in this case, the coolant does not move in the energy absorbing space (EAS) and the energy transfer space (ETS). FIG. 4B illustrates the movement of the coolant in the state in which the reactor stops due to accidents, etc., and thus starts to release overheat. First, as illustrated in FIG. 4B, when the pressure within the energy release space (ERS) is increased due to the overheat of the reactor driving system (in this case, pressurization may be made only by air within the energy release space (ERS) or when the coolant is leaked in the reactor driving system, more pressurization may be made by steam generated by the evaporation of the leaked coolant), the coolant within the energy absorbing space (EAS) is pressurized through the passage formed over the energy release space (ERS). Therefore, as illustrated by an arrow, the coolant within the energy absorbing space (EAS) is supplied to the energy transfer space (ETS). Meanwhile, in the heat exchange device within the energy transfer space (ETS), a heat exchanger which is connected to the reactor driving system within the energy release space (ERS) and a heat exchanger which is provided within the energy transfer space (ETS) are adjacently provided to each other to exchange heat therebetween. Further, the coolant within the energy absorbing space (EAS) is supplied to the heat exchange device, such that the heat exchange in the heat exchange device is faster. (This is based on the two-phase heat transfer principle which will be described below in more detail) That is, in the case of FIG. 4B, indirect cooling of the reactor driving system is performed while the coolant moves from the energy absorbing space (EAS) to the energy transfer space (ETS). FIG. 4C illustrates the movement of the coolant in the state in which overheat is more generated and thus sufficient cooling may not be made only by the indirect cooling as illustrated in FIG. 4B. As illustrated in FIG. 4B, if the cooling of the reactor driving system is continued to be performed while the coolant moves from the energy absorbing space (EAS) to the energy transfer space (ETS), a level of the coolant within the energy transfer space (ETS) continuously rises over time and a level of the coolant within the energy absorbing space (EAS) continuously falls. In this case, when the level of the coolant within the energy transfer space (ETS) rises beyond a certain level, the coolant within the energy transfer space (ETS) moves to the energy release space (ERS) as illustrated in FIG. 4C. The coolant moving to the energy release space (ERS) directly contacts the reactor driving system, such that the coolant may directly absorb heat from the reactor driving system. That is, in the case of FIG. 4C, the direct cooling of the reactor driving system is performed while the coolant moves from the energy transfer space (ETS) to the energy release space (ERS). As described above, in the case of FIGS. 4B and 4C, the heat released from the reactor driving system is transferred to the coolant, such that the cooling of the reactor driving system may be performed. As such, according to the exemplary embodiment of the present invention, the cooling of the reactor driving system is performed by appropriately moving the coolant within the reactor safety system between the respective spaces depending on the thermal-hydraulic conditions such as the change in pressure in the reactor driving system and whether the coolant is leaked, and the change in pressure in the respective spaces and the level of the coolant which are changed depending thereon. Here, in the reactor according to the exemplary embodiment of the present invention, a heat transfer is performed using, in particular, a two-phase heat transfer mechanism. FIG. 5 is a diagram for describing a principle of the two-phase heat transfer mechanism which is used in the reactor according to the exemplary embodiment of the present invention. A basic heat exchange principle in most of the existing heat exchangers is to pass a heat exchange medium through a channel isolated from the outside and directly exchange heat between a heat exchange medium in the channel and another heat exchange medium outside the channel, having a wall surface of the channel provided therebetween, thereby performing the heat transfer. That is, for example, an air cooling type heat exchanger form which is configured to make high-temperature cold water flow inside the channel and make low-temperature air flow outside the channel to allow the low-temperature air to absorb the heat of the high-temperature cold water has been most widely used. As another example, a heterogeneous heat exchanger form which is configured to make high-temperature oil flow in one side thereof and the low-temperature cold water flow in the other side thereof to allow the low-temperature cool to absorb the heat of the high-temperature oil has been widely used as well. The heat exchange device according to the exemplary embodiment of the present invention exchanges heat depending on an entirely different principle therefrom. As illustrated in FIG. 5, the heat exchange device according to the exemplary embodiment of the present invention using the two-phase heat transfer mechanism basically includes a releasing tube (left tube in FIG. 5) in which the high-temperature heat exchange medium flows, an absorbing tube (right tube in FIG. 5) in which the low-temperature heat exchange medium flows, and a nozzle through which other heat exchange media (cold water in FIG. 5 or other liquids) are sprayed to the two tubes. The high-temperature heat exchange medium flows in the releasing tube and the low-temperature heat exchange medium flows in the absorbing tube. In the case of the existing heat exchanger, the two tubes adhere to each other to transfer heat from a high temperature side to a low temperature side through a wall surface of the tube. However, in the heat exchange device according to the exemplary embodiment of the present invention using the two-phase heat transfer mechanism, unlike the configuration of the two tubes of the existing heat exchanger, the two tubes are spaced apart from each other at an appropriate interval. The nozzle is provided at the releasing tube side to spray the cold water into the releasing tube. If the cold water is sprayed and water drops thereof approach or contact an outer surface of the releasing tube, the water drops of the cold water instantly absorb heat of the high-temperature heat exchange medium within the releasing tube and thus are quickly evaporated. That is, the water drops of the cold water quickly absorb a large amount of evaporation heat at an outer surface of the releasing tube and thus tube outside quenching happens and the high-temperature heat exchange medium is released while the heat thereof being taken away as the evaporation heat of the water drops of the cold water inside the releasing tube and thus tube inside condensation happens. As described above, the cold water around the releasing tube is completely evaporated and becomes a steam state. The steam contacts the absorbing tube which is spaced apart from the releasing tube. In this case, since the low-temperature heat exchange medium flows in the absorbing tube, if the steam approaches or contacts the outer surface of the absorbing tube, the heat of the steam is instantly taken away to the low-temperature heat exchange medium within the absorbing tube and thus the steam is condensed, such that the steam is formed at the outer surface of the absorbing tube. That is, since the heat of the steam is taken away to the low-temperature heat exchange medium at the outer surface of the absorbing tube, the steam is condensed and thus becomes condensed water, such that tube outside condensing happens and the low-temperature heat exchange medium absorbs heat from the steam inside the absorbing tube, such that tube inside evaporation happens. As such, in the two-phase heat transfer mechanism, the heat exchange medium (cold water in an example of FIG. 5) sprayed from the nozzle performs the heat transfer while being changed to gas phase—liquid phase in such a manner that the releasing tube and the absorbing tube are spaced apart from each other and the heat exchange medium is sprayed in a liquid phase, evaporated near the releasing tube to be in a gas phase, and condensed near the absorbing tube and again returns to the liquid phase. A research result of the two-phase heat transfer scheme to transfer heat even faster and more effectively than the existing heat transfer scheme has been published recently. The safety system according to the exemplary embodiment of the present invention uses the two-phase heat transfer scheme to allow the coolant to absorb heat, thereby realizing the cooling faster and more efficiently than the existing reactor safety system. The conceptual configuration and principle of the present invention has been described with reference to FIGS. 3 to 5. Hereinafter, the reactor according to the exemplary embodiment of the present invention will be described below in more detail. FIG. 6 is a diagram illustrating the reactor according to the exemplary embodiment of the present invention. As illustrated in FIG. 6, the reactor according to the exemplary embodiment of the present invention is also configured to include the reactor driving system and the reactor safety system. The reactor driving system is configured to include a reactor vessel 112 accommodating a reactor core 111 and a steam generator 113 to which a steam pipe 114 and a water supply pipe 115 are connected. As described above, the steam generator may be integrally formed within the reactor vessel or the steam generator may be provided outside the reactor vessel. Therefore, the reactor driving system is not limited to the illustrated form but may be formed in any form. An operation of each part of the reactor driving system will be described below in more detail. The reactor core 111 is a core part of the reactor and is a part of causing the nuclear fission in which an atomic nucleus of nuclear fuel is split into two by being combined with a neutron to generate heat energy. That is, generally, the reactor core 111 is called a bundle of fuel rods which is nuclear fuel of the reactor. Further, generally, the reactor core 111 includes a reactor output control rod, in which the reactor output control rod is vertically movably inserted into the reactor core 111 to control how much the nuclear fission of the nuclear fuel is made depending on the inserted degree of the reactor output control rod, and thus acts to control the output of the reactor 100. The reactor vessel 112 accommodates the reactor core 111 while being sealed off from the outside. As described above, since the reactor output control rod needs to be controlled to vertically move, a portion of the upper end of the reactor output control rod is generally provided to be exposed outside the reactor vessel 112. In this respect, the reactor core 111 is naturally disposed under the reactor vessel 112. In this case, the coolant is accommodated in the reactor vessel 112 and thus the heat energy generated from the reactor core 111 is absorbed into the coolant. The coolant absorbs the heat energy generated from the reactor core 111 to act to cool the reactor core 111 and transfers the heat absorbed by the coolant to the outside, thereby generating power (hereinafter, which will be described below in more detail in the steam generator 113). As the coolant, cold water, that is, water is generally used. Although described below in more detail, the heat exchange media distributed in each part may be mixed due to the opening of a circulation channel which is in a closed state at ordinary times, etc., when the reactor safety system is operated, and therefore the heat exchange medium or the coolant which is used in each part of the reactor is generally used as the cold water. The steam generator 113 is formed in the heat exchanger form and is provided inside the reactor vessel 112. The coolant operated as the heat exchange medium is distributed inside the steam generator 113 and the steam generator 113 is supplied with heat from the coolant inside the reactor vessel 112 around the steam generator 113. Therefore, the coolant distributed inside the steam generator 113 absorbs heat and the evaporation happens. As such, the coolant which is in the high-temperature, high-pressure gas phase is released to the steam pipe 114 to operate a turbine. After the turbine is operated, the condensed coolant is again supplied to the steam generator 113 through the water supply pipe 115 and thus is circulated. The steam pipe 114 and the water supply pipe 115 are each provided with a steam pipe isolation valve 114a and a water supply pipe isolation valve 115a and thus are blocked from the outside at the time of an emergency situation. As such, when the reactor is normally operated, the coolant inside the reactor vessel 112 is naturally circulated. This will be described below in more detail. When the heat energy generated from the reactor core 111 is absorbed into the coolant, the high-temperature coolant rises. When the high-temperature coolant reaches the steam generator 113 which is disposed over the reactor core 111, the coolant within the steam generator 113 and the high-temperature coolant exchange heat with each other. That is, the coolant within the steam generator 113 absorbs heat from the high-temperature coolant. Therefore, the high-temperature coolant has the reduced temperature while passing through the steam generator 113, and thus falls. The falling coolant again absorbs the heat energy generated from the reactor core 111, such that the natural circulation convection is made. The foregoing reactor driving system may be commonly included in the existing large or small (integrated) reactor. The driving system configured as described above is operated and thus the reactor produces power. When the reactor is normally operated, the operation of the reactor has no problem only by the driving system. However, if the reactor vessel 112 is damaged and thus the coolant within the reactor vessel 112 is leaked, the heat energy generated from the reactor core 111 may not be absorbed into a sufficient amount of coolant. Therefore, the temperature around the reactor core 111 excessively rises and thus more serious damage such as melting of parts may occur. The reactor has a great effect on environment such as radioactivity leakage and therefore safety is important above all. Therefore, if the coolant is leaked due to the damage of the reactor vessel 112, a safety system to quickly cool the reactor vessel 112, etc., is essential. As described above, if the safety system is operated only when receiving a separate control command such as an operator's manipulation, the operator of the reactor does not issue a command in time at the time of the occurrence of accidents under the conditions that he/she is absent, wounded, or the like. Therefore, the danger of the accident is terribly increased. Further, even in the case of the system which is automatically controlled by an electron control, etc., when the safety system is damaged due to the high heat generated due to the damage of the reactor, the safety system may not be correctly operated. Therefore, when the reactor is damaged and the coolant is leaked, a passive safety system which is mechanically operated upon the change in physical environment is necessarily provided. Of course, many studies on the passive safety system have been conducted before. However, the existing passive safety system has the following problems. First, since the existing passive safety systems have a considerably large volume, there are many restrictions in construction and economy. As a result, considerable resources such as time and manpower are consumed to design the safety systems for overcoming the above problems. In particular, in the case of the small reactor of which the demand has been gradually expanded recently, if the existing passive safety system is applied to the small reactor, the small reactor may not obtain the sufficient cooling efficiency and thus the difficulty in designing the small reactor may be increased. Therefore, resources such as a space, a material, and costs may be more wasted during the actual construction of the reactor and resources such as time, manpower, and costs consumed for maintenance, etc., cannot but be more wasted even during the operation of the constructed reactor. To solve the above problems, the exemplary embodiment of the present invention may appropriately cope with the various occurrence situations of accidents of reactors using one system and may configure the safety systems (that is, devices for cooling a reactor) in completely passive type and thus there is no need for the separate control command of the operator. As a result, the fast cooling may be made when the reactor is damaged. In particular, unlike the existing reactor, the reactor according to the exemplary embodiment of the present invention mostly uses the two-phase heat transfer scheme to perform the cooling and performs the cooling while the coolant appropriately moves within the containment vessel divided into several spaces at the time of the occurrence of accidents to perform the faster and more efficient cooling than before. In addition, the safety system according to the exemplary embodiment of the present invention has a configuration much simpler than that of the existing passive safety system to greatly reduce the volume of the reactor, such that it is very suitable for the small reactor of which the demand has been gradually expanded recently. In FIG. 6, the reactor safety system is configured to include a releasing isolation vessel 120, a releasing heat exchanger 121, a releasing heat exchange channel 122, an absorbing isolation vessel 130, an absorbing heat exchanger 131, an absorbing heat exchange channel 132, a condensing heat exchanger, a coolant spray pipe 134, and a transferring isolation vessel 140. In this configuration, the releasing isolation vessel 120 forms the energy release space (ERS) in the conceptual configuration of FIG. 3, the absorbing isolation vessel 130 forms the energy absorbing space (EAS) in the conceptual configuration of FIG. 3, and the transferring isolation vessel 140 forms the energy transfer space (ETS) in the conceptual configuration of FIG. 3. Further, the releasing heat exchanger 121, the absorbing heat exchanger 131, and the coolant spray pipe 134 are combined to form the heat exchange device in the conceptual configuration of FIG. 3. The indirect cooling based on the scheme as illustrated in FIG. 4B may be performed only by the above configuration. In addition, to perform the direct cooling based on the scheme as illustrated in FIG. 4C, the reactor safety system is configured to further include the coolant injection pipe 135. Further, to more smoothly move the coolant between the respective components of the reactor safety system, the reactor safety system may further include several additional components. Hereinafter, the connection relationship between the respective components will be first described. Next, how the respective components are operated in some cases will be described in detail with reference to the embodiment of the indirect cooling operating method of FIG. 7 and the embodiment of the direct cooling operating method of FIG. 8. The releasing isolation vessel 120 accommodates gas and the reactor driving system as illustrated and forms the energy release space (ERS) as described above. The absorbing isolation vessel 130 accommodates the coolant as illustrated and forms the energy absorbing space (EAS) as described above. The absorbing isolation vessel 130 communicates with the releasing isolation vessel 120 through a passage 123 which is formed over the releasing isolation vessel 120. In this case, as illustrated in FIG. 6, the passage 123 may be formed in a double barrier form in which a releasing isolation vessel barrier 123a is formed at the releasing isolation vessel 120 and has an upper portion opened and an absorbing isolation vessel barrier 123b is formed at the absorbing isolation vessel 130 and has a lower portion opened. By the configuration, it is possible to prevent the coolant within the absorbing isolation vessel 130 from unnecessarily passing to the releasing isolation vessel 120 and freely distribute air or steam. The transferring isolation vessel 140 is provided over the absorbing isolation vessel 130 as illustrated to accommodate the gas and the coolant and forms the energy transfer space (ETS) as described above. The transferring isolation vessel 140 includes the heat exchange device (hereinafter, the heat exchange device including the releasing heat exchanger 121, the absorbing heat exchanger 132, and the coolant spray pipe 134 which will be described below) for transferring heat released from the reactor driving system to the coolant and the heat exchange device uses the two-phase heat transfer mechanism as described above to transfer heat. In this case, the condensed water, that is, the condensed coolant essentially falls down the heat exchange device. To prevent the coolant from being unnecessarily introduced into the energy release space (ERS), as illustrated in FIG. 6, the inside of the transferring isolation vessel 140 is provided with an accommodating barrier 141 enclosing an area in which the releasing heat exchanger 121 and the absorbing heat exchanger 131 are disposed, such that the transferring isolation vessel 140 may accommodate the cold water. Further, as illustrated in FIG. 3 or 6, the energy absorbing space (EAS) and the energy transfer space (ETS) may be preferably formed in a form in which one space is vertically divided. Further, in the direct cooling scheme which is conceptually described in FIG. 4B, the coolant filled in the energy transfer space (ETS) is directly introduced into the energy release space (ERS). In this case, as the energy transfer space (ETS) is positioned high, the coolant may flow smoother and smoother, and therefore the position of the energy transfer space (ETS) may be preferably formed higher than the position of the energy release space (ERS) as commonly illustrated in FIG. 3 or 6. The releasing heat exchanger 121 and the absorbing heat exchanger 131 are adjacently disposed to each other within the transferring isolation vessel 140 to exchange heat with each other. The coolant spray pipe 134 is configured to inject the coolant into the releasing heat exchanger 121 and the absorbing heat exchanger 131 which are adjacently disposed to each other, thereby forming the heat exchange device in the conceptual configuration of FIG. 3. By this configuration, the coolant sprayed from the coolant spray pipe 134 absorbs heat from the coolant distributed within the releasing heat exchanger 121 to be evaporated and the coolant distributed within the absorbing heat exchanger 131 absorbs the heat of the steam generated by the evaporation of the coolant to be condensed and is formed to be accommodated into the transferring isolation vessel 140, such that the heat transfer may be made by the two-phase heat transfer mechanism for transferring heat from the coolant within the releasing heat exchanger 121 to the coolant within the absorbing heat exchanger 131 by the evaporation and the condensing of the coolant sprayed by the coolant spray pipe 134. As such, the heat exchange device configured to include the releasing heat exchanger 121, the absorbing heat exchanger 131, and the coolant spray pipe 134 is a connection component for absorbing the heat of the reactor driving system and discarding the absorbed heat to the coolant. As illustrated in FIG. 6, the heat exchange device further includes a condensing heat exchanger 133 disposed in the external environment of the isolation vessels, the releasing heat exchange channel 122 connected to the reactor vessel 112 and the releasing heat exchanger 121 to circulate and distribute the coolant, and the absorbing heat exchange channel 132 connected to the absorbing heat exchanger 131 and the condensing heat exchanger 133 to circulate and distribute the coolant. Meanwhile, only if the reactor safety system performs the cooling only by the heat transfer using the two-phase heat transfer mechanism, that is, if the cooling may be made only by the indirect cooling scheme of FIG. 4B described above, the coolant spray pipe 134 may be supplied with the coolant from other places. In this case, the difficulties in design happen. Therefore, to rule out the difficulties in design and use the direct cooling scheme of FIG. 4C, the coolant spray pipe 134 has one end communicating with the absorbing isolation vessel 130 to be supplied with the coolant and the other end provided with the nozzle to spray the supplied coolant to the releasing heat exchanger 121 and the absorbing heat exchanger 131 and is preferably provided with a coolant spray valve 134a. Further, for the reactor safety system to perform the cooling using the direct cooling scheme of FIG. 4C, a passage through which the coolant is released from the energy transfer space (ETS) to the energy release space (ERS) is required. To this end, the reactor safety system further includes a coolant injection pipe 135 which has one end communicating with the transferring isolation vessel 140 and the other end communicating with the releasing isolation vessel 120 to inject the coolant accommodated in the transferring isolation vessel 140 into the releasing isolation vessel 120 and a coolant injection valve 135a provided on the coolant injection pipe 135. Hereinafter, how the respective components are operated in some cases will be described in detail with reference to the indirect cooling operating method of FIG. 7 and the direct cooling operating method of FIG. 8. Further, added components will be described while an auxiliary operation process is described during the operation. In operating the reactor configured as described above, first, the indirect cooling scheme as schematically described in FIG. 4B may be used. FIG. 7 illustrates an example of the operation state in this case. First, starting the operation of the reactor safety system means that the reactor driving system is abnormally operated due to the intentional stop of the reactor driving system, the accidents of the reactor, etc. That is, prior to starting the operation of the reactor safety system, isolating which includes stopping the reactor driving system; and closing the steam pipe isolation valve 114a included in the steam pipe 114 and the water supply pipe isolation valve 115a included in the water supply pipe 115 is always performed. In FIG. 7, the actual operation of the reactor safety system includes circulating the releasing heat exchanger, performing the two-phase heat transfer, and circulating the absorbing heat exchanger. The circulating of the releasing heat exchanger is performed in the following order. First, the steam generated by the evaporation of the coolant within the reactor vessel 112 is introduced into the releasing heat exchanger 121 through the releasing heat exchange channel 122. Next, the steam introduced into the releasing heat exchanger 121 is condensed while passing through the releasing heat exchanger 121 to generate the coolant and the generated coolant is reintroduced into the reactor vessel 112 through the releasing heat exchange channel 122 to be circulated. The performing of the two-phase heat transfer is performed in the following order. First, the coolant supplied by the coolant spray pipe 134 is sprayed to and contacts the outer surface of the releasing heat exchanger 121. Next, the coolant contacting the outer surface of the releasing heat exchanger 121 absorbs the heat from the steam within the releasing heat exchanger 121 to be evaporated and the steam within the releasing heat exchanger 121 is condensed to generate the coolant. As such, if the steam generated by the evaporation of the coolant at the outer surface of the releasing heat exchanger 121 contacts the outer surface of the absorbing heat exchanger 131, the steam contacting the outer surface of the absorbing heat exchanger 131 releases heat to the coolant within the absorbing heat exchanger 131 and thus is condensed to the coolant and the coolant within the absorbing heat exchanger 131 is evaporated to generate the steam. The circulating of the absorbing heat exchanger is performed in the following order. First, the steam within the absorbing heat exchanger 131 is introduced into the condensing heat exchanger 133 through the absorbing heat exchange channel 132. Next, the steam introduced into the condensing heat exchanger 133 is condensed while passing through the condensing heat exchanger 133 to generate the coolant and the generated coolant is reintroduced into the absorbing heat exchanger 131 through the absorbing heat exchange channel 132 to be circulated. That is, the heat of the reactor driving system moves to the releasing heat exchanger by the circulating of the releasing heat exchanger, the heat is exchanged between the releasing heat exchanger and the absorbing heat exchanger by the performing of the two-phase heat transfer to transfer heat to the coolant, the heat of the heat exchange medium (in most cases, made of the same material as the coolant) within the absorbing heat exchanger is discarded to the external environment by the circulating of the absorbing heat exchanger, such that the indirect cooling of the reactor driving system may be performed. In this case, as illustrated, the condensing heat exchanger 133 is included in the heat sink having infinite capacity such as sea water, thereby effectively cooling the heat of the reactor driving system. Meanwhile, when the cooling is performed by the indirect cooling scheme, reactor vessel 112-releasing heat exchange channel 122-releasing heat exchanger 121 form one isolation circulation space and absorbing heat exchanger 131-absorbing heat exchange channel 132-condensing heat exchanger 133 forms another isolation circulation space. In this case, the isolating is first performed, such that steam pipe 114-steam generator 113-water supply pipe 115 forms another isolation space. However, the isolation spaces in which the releasing heat exchanger or the absorbing heat exchanger is included are configured to have the coolant circulated therein, while the isolation space in which the steam generator is included is merely isolated. In this case, the isolation space in which the steam generator is included is a space which is connected to the external turbine to have the coolant and the steam distributed therein at the time of the normal operation, and therefore the coolant which is confined in the space at the time of the isolation remains. The coolant is also evaporated within the isolation space due to the overheat of the reactor core to be changed to the steam. Here, the steam excessively increases the pressure within the isolation space in which the steam generator is included, and thus the steam generator, etc., may be damaged. To prevent the problems, the reactor safety system further includes a steam bypass pipe 126 having one end communicating with the steam pipe 114 and the other end communicating with the releasing heat exchange channel 122 to distribute the steam generated by the evaporation of the coolant within the steam generator 113 to the releasing heat exchange channel 122 and a steam bypass valve 126a provided on the steam bypass pipe 126, and the operating method for the reactor may further include the following steps. That is, the operating method for the reactor may further include: pressurizing the space within the steam generator 113 by the steam generated by the evaporation of the coolant within the steam generator 113; and opening the steam bypass valve 126a by the pressure to introduce the steam within the steam generator 113 into the releasing heat exchange channel 122 through the steam bypass pipe 126, such that the steam auxiliary introduction into the releasing heat exchange channel 122 may be additionally performed. By doing so, the steam auxiliary introduction is made, and at the same time it is possible to prevent the pressure within the isolation space including the steam generator 113 from excessively increasing. Meanwhile, the steam released from the isolation space including the steam generator 113 is condensed in the coolant state while passing through the releasing heat exchanger 121. The coolant returns to the reactor vessel 112 through the releasing heat exchange channel 122 to be supplemented within the reactor vessel 112. However, if the steam released from the space of the steam generator is continued to be merely introduced into the reactor vessel, the steam within the space of the steam generator and the amount of the coolant are insufficient with the passage of time and thus the pressure may be excessively reduced, which makes the state of the reactor instable. Therefore, if the overpressure within the steam generator 113 falls to a proper level, it is preferable to re-supply the coolant to the isolation space including the steam generator 113. To this end, the reactor safety system further includes a water supply supplement pipe 128 having one portion connected to the releasing heat exchange channel 122 and the other portion connected to the water supply pipe 115 to supply the coolant, which is introduced through the releasing heat exchange channel 122, into the steam generator 112. As illustrated, the water supply supplement pipe 128 includes a water supply supplement valve 128a. Further, a valve is also provided between a portion where the condensed coolant flows in the reactor vessel 122 from the releasing heat exchange channel 122 connected to the reactor vessel 122 and a portion to which water supply supplement pipe 128 is connected and valves are opened and closed depending on a proper pressure condition, thereby easily realizing the foregoing operation. If the coolant is not leaked in the reactor driving system, that is, if the reactor is not damaged but is in the overheat state due to the slight abnormal operation, it is possible to prevent the overheat damaging the reactor from proceeding only by the cooling of the reactor driving system using the indirect cooling scheme as described above. However, when a loss-of-coolant accident (LOCA) (the accident that the coolant is leaked from the reactor vessel) or a medium and large accident that the reactor core is melted and then is exposed through the reactor vessel occurs, there is a need to perform the cooling faster and more efficiently. In this case, the direct cooling scheme as schematically described in FIG. 4C may be used. FIG. 8 illustrates an example of the operation state in this case. The exemplary embodiment of FIG. 8 is performed under the assumption that the exemplary embodiment of FIG. 7, that is, the indirect cooling scheme is first performed. In other words, the exemplary embodiment of FIG. 8 is performed under the assumption that the steps performed in the exemplary embodiment of FIG. 7, that is, the steps of distributing the coolant into the isolation circulation space including the releasing heat exchanger or the absorbing heat exchanger are performed, are performed in advance, or the like. In the exemplary embodiment of FIG. 8, the operation of the reactor safety system includes pressurizing the releasing isolation vessel 120 and spraying the coolant, and directly cooling by the coolant. In this case, to perform the exemplary embodiment of FIG. 8, the coolant spray pipe 134 draws the coolant from the absorbing isolation vessel 130 and sprays the drawn coolant. Further, a component to release the coolant from the absorbing isolation vessel 130 to the releasing isolation vessel 120 is additionally required. Describing in more detail, to perform the exemplary embodiment of FIG. 8, first, the coolant spray pipe 134 has one end communicating with the absorbing isolation vessel 130 to be supplied with the coolant and the other end provided with the nozzle to spray the supplied coolant to the releasing heat exchanger 121 and the absorbing heat exchanger 131 and has the coolant spray valve 134a provided thereon. Further, the reactor safety system further includes a coolant injection pipe 135 which has one end communicating with the transferring isolation vessel 140 and the other end communicating with the releasing isolation vessel 120 to inject the coolant accommodated in the transferring isolation vessel 140 into the releasing isolation vessel 120 and a coolant injection valve 135a provided on the coolant injection pipe 135. The pressurizing the releasing isolation vessel 120 and spraying of the coolant is performed in the following order. First, the steam generated by the evaporation of the coolant within the reactor vessel 112 is leaked and is filled in the space within the releasing isolation vessel 120 to pressurize the releasing isolation vessel 120. Next, the pressure is transferred to the coolant within the absorbing isolation vessel 130 through the path 123 and is thus introduced into the coolant spray pipe 134. Therefore, the coolant spray valve 134a is opened by the pressure, and thus the coolant is sprayed to the releasing heat exchanger 121 and the absorbing heat exchanger 131 through the coolant spray pipe 134, such that the coolant within the absorbing isolation vessel 130 may move to the transferring isolation vessel 140 to be supplied to the heat exchangers. The directly cooling by the coolant is performed in the following order. As described above, it is on the assumption that the steps in the exemplary embodiment of FIG. 7 proceed in advance, such that the coolant generated by condensing steam by performing the two-phase heat transfer is accommodated within the transferring isolation vessel 140. If the accommodating barrier 141 is formed around the heat exchangers, even though the coolant generated by condensing steam is filled in the transferring isolation vessel 140, the coolant is filled in the accommodating barrier 141 and therefore is not directly introduced into the coolant injection pipe 135. However, after some time lapses, the amount of coolant is increased and thus the level of the coolant is increased, such that the coolant accommodated within the transferring isolation vessel 140 is introduced into the coolant injection pipe 135. In this case, the higher the level of the coolant within the transferring isolation vessel 140, the smaller the size of the empty space, that is, the space filled with gas, such that the pressure within the transferring isolation vessel 140 is increased. Therefore, the coolant injection valve 135a is opened by the pressure and thus the coolant is injected into the releasing isolation vessel 120 through the coolant injection pipe 135. As such, the coolant injected into the releasing isolation vessel 120 directly contacts the reactor vessel 112 to perform the cooling. In this case, as described above, the pressure within the transferring isolation vessel 140 needs to be increased to a certain level to open the coolant injection valve 135a. Therefore, in order to open the coolant injection valve 135a faster, the following components may further be added. That is, the reactor safety system may further include a releasing isolation vessel communicating pipe 124 formed to communicate between the upper portion of the releasing isolation vessel 120 and the upper portion of the transferring isolation vessel 140 and a releasing isolation vessel communicating valve 124a provided on the releasing isolation vessel communicating pipe 124. Further, the operating method for the reactor further includes: leaking the steam generated by the evaporation of the coolant within the reactor vessel 112 to fill the space within the releasing isolation vessel 120 so as to pressurize the releasing isolation vessel 120; and opening the releasing isolation vessel communicating valve 124a by the pressure to release the steam within the releasing isolation vessel 120 to the transferring isolation vessel 140 through the releasing isolation vessel communicating pipe 124 so as to fill and pressurize the transferring isolation vessel 140, such that the auxiliary pressurization for opening the coolant injection valve 135a in the direct cooling by the coolant may be additionally performed. Meanwhile, when the middle and large accident occurs, enormous heat is released within the releasing isolation vessel 120 accommodating the reactor driving system. Therefore, the coolant within the releasing isolation vessel 120 is almost all evaporated at the early stage that the cooling is little performed and thus becomes the steam state, such that the pressure within the releasing isolation vessel 120 is increased. In addition, if the coolant is continued to be injected from the transferring isolation vessel 140, the pressure within the releasing isolation vessel 120 may be excessively increased. To prevent the above problem, the reactor safety system may further include the following components. That is, the reactor safety system may further include a releasing isolation vessel pressure reducing pipe 125 formed to communicate between a lower portion of the releasing isolation vessel 120 and a lower portion of the absorbing isolation vessel 130 and a releasing isolation vessel pressure reducing valve 125a provided on the releasing isolation vessel pressure reducing pipe 125. Further, the operating method for the reactor further includes: leaking the steam generated by the evaporation of the coolant within the reactor vessel 112 to fill the space within the releasing isolation vessel 120 so as to pressurize the releasing isolation vessel 120; filling the releasing isolation vessel 120 with the coolant by the direct cooling by the coolant; and opening the releasing isolation vessel pressure reducing valve 125a by the pressure to release the coolant within the releasing isolation vessel 120 to the absorbing isolation vessel 130 through the releasing isolation vessel pressure reducing pipe 125, such that the pressure reduction of the overpressure formed within the releasing isolation vessel 120 may be additionally performed. Meanwhile, all the operations of the safety systems are performed by increasing the pressure within the reactor vessel 112 or the releasing isolation vessel 120. That is, the safety operation for cooling may be performed faster by increasing the pressure within the releasing isolation vessel 120 faster. To this end, the reactor safety system may further include the following components. That is, the reactor safety system may further include a steam release pipe 127 having one end communicating with the steam pipe 114 and the other end communicating with the space within the releasing isolation vessel 120 to release the steam generated by the evaporation of the coolant within the reactor vessel 112 to the space within the releasing isolation vessel 120 and a steam release valve 127a provided on the steam release pipe. Further, the operating method for the reactor may further include: pressurizing the space within the steam generator 113 by the steam generated by the evaporation of the coolant within the steam generator 113; and opening the steam release valve 127a by the pressure to release the steam within the steam generator 113 to the space within the releasing isolation vessel 120 through the steam release pipe, such that the auxiliary pressurization within the releasing isolation vessel 120 may be additionally performed. Further, as described above, the reactor safety system according to the exemplary embodiment of the present invention performs the cooling basically using the two-phase heat transfer mechanism. In this case, the coolant and the steam which are originally present in the reactor vessel 112 are circulated within the isolation space including the releasing heat exchanger 131. However, when the coolant is leaked from the reactor vessel 112, the amount of coolant circulated to be used for the cooling may be insufficient. To prevent the above problem, the reactor safety system may further include the following components. That is, the reactor safety system may further include a coolant supplement pipe 136 having one end communicating with the releasing heat exchange channel 122 and the other end communicating with a space under the surface of the coolant within the absorbing isolation vessel 130 or the transferring isolation vessel 140 to supplement the coolant to the releasing heat exchange channel 122 and a coolant supplement valve 136a provided on the coolant supplement pipe 136. Further, the operating method for the reactor further includes opening the coolant supplement valve 136a by the pressure to introduce the coolant into the releasing heat exchange channel 122 through the coolant supplement pipe 136, such that the auxiliary introduction of the coolant into the releasing heat exchange channel 122 may be additionally performed. As described above, in the reactor safety system according to the exemplary embodiment of the present invention, the coolant performs cooling while moving among the releasing isolation vessel 120, the absorbing isolation vessel 130, and the transferring isolation vessel 140. During the process, the condensing heat exchanger 133 connected to the absorbing heat exchanger 131 is included in the heat sink having infinite capacity like sea water to extend the cooling time infinitely, thereby performing the long-term cooling. However, as described above, the heat is directly absorbed by the coolant circularly moving within the isolation vessels and heat continuously discarded is accumulated in the coolant and thus the temperature of the coolant rises, thereby reducing the cooling speed and performance. Therefore, the reactor safety system according to the exemplary embodiment of the present invention further includes a structure to auxiliarily further cool the coolant within the isolation vessels. The auxiliary waste heat removing part 150 performs this role and the auxiliary waste heat removing part 150 includes an auxiliary absorbing heat exchanger 151 disposed within the absorbing isolation vessel 130, an auxiliary condensing heat exchanger 153 disposed outside the isolation vessels, and an auxiliary absorbing heat exchange channel 152 connected to the auxiliary absorbing heat exchanger 151 and the auxiliary condensing heat exchanger 153 to circularly distribute the coolant. In this case, the operating method for the reactor may further include operating the auxiliary waste heat removing part 150, that is, removing the auxiliary waste heat. The removing of the auxiliary waste heat is a substantially completely independent operation and therefore the removing of the auxiliary waste heat may be performed in parallel with all the steps of the operating method for the reactor safety system described above. The removing of the auxiliary waste heat is performed as follows. First, the coolant within the auxiliary absorbing heat exchanger 151 absorbs heat from the coolant within the absorbing isolation vessel 130. Next, the coolant within the auxiliary absorbing heat exchanger 151 is introduced into the auxiliary condensing heat exchanger 153 through the auxiliary absorbing heat exchange channel 152. Next, the coolant within the auxiliary condensing heat exchanger 153 releases heat to the external environment. In this case, the external environment may be sea water similar to the external environment of the condensing heat exchanger 133 or may be an underground layer, etc., as illustrated. Next, the coolant within the auxiliary condensing heat exchanger 153 is circulated by being reintroduced into the auxiliary absorbing heat exchanger 151 through the auxiliary absorbing heat exchange channel 152. According to the exemplary embodiments of the present invention, the safety system is completely passively performed to perform the fast cooling without the separate control command at the time of the damage of the reactor, thereby minimizing the accident risk. First of all, the exemplary embodiment of the present invention may have the new structure based on the thermal-hydraulic operation concept completely different from the existing passive reactor safety system and variously perform the appropriate cooling operation to meet various accident conditions based on the new structure. In particular, the configuration of the reactor safety system according to the exemplary embodiment of the present invention has a much simpler structure than the typical reactor safety system, such that it is possible to much reduce the difficulty in designing and constructing the reactor than the existing reactor and it is possible to improve the easiness and convenience beyond compare even in operating the constructed reactor. As a result, it is possible to greatly save the resources such as time, manpower, and costs consumed for the design, the construction, the operation, the control, etc. Further, the configuration of the reactor safety system according to the exemplary embodiment of the present invention is different from that of the existing reactor safety system and quickly absorbs the high-temperature energy of the reactor using the two-phase heat transfer scheme unlike the existing reactor safety system to discard the high-temperature energy to the outside (heat sink such as sea water) so as to make the cooling speed remarkably fast, thereby greatly improving the absolute cooling ability. Further, the heat absorbed into the coolant acting as the heat sink is discarded to the outside, that is, the infinite heat sink once more, such that the residual heat removal time may be extended infinitely. In addition, there is no active control means (operated by receiving the control command issued by the operator from the outside) in performing the cooling operation and the general structure itself is much simpler, thereby more improving the convenience of manufacturing, operation, etc, than before. Further, the economical effect of saving the unnecessary energy waste during the operation may be achieved since the separate power source required for the active control or driving is not required. In addition, according to the exemplary embodiment of the present invention, the shape of the reactor system is more simpler than before and (since the cooling ability is more excellent than before) the reactor may be installed in the space much smaller than the reactor including the existing safety system, thereby much improving the operation and construction economic efficiency of the reactor. The present invention is not limited to the above-mentioned embodiments but may be variously applied, and may be variously modified by those skilled in the art to which the present invention pertains without departing from the gist of the present invention claimed in the claims.
	abstract	An electron beam lithographing method for exposing and lithographing a desired pattern with an electron beam includes the steps of forming a plurality of accuracy evaluation patterns and a desired pattern in a stripe connection boundary area so as to form an electron beam mask, and measuring connection errors of the exposed stripes with the accuracy evaluation patterns.
059986891	abstract	A method for recycling metal parts contaminated by radioactive elements, in particular by .alpha.-emitters, includes forming a melt and a slag from the metal parts and then separating the slag from the melt. The radioactive elements are oxidized prior to the formation of the melt and the slag. For that purpose, the contaminated metal parts are exposed to an oxygen-containing atmosphere for a period at a temperature below the melting temperature of the metal parts.
	summary
	claims	1. A fluoroscopic imaging method performed during a dental procedure for providing real-time video display of the dental procedure, comprising the steps of:(a) causing a beam of electromagnetic radiation to travel,(1) from an emitter,(2) through an area in which the dental procedure is being performed on a patient,(3) to a flat panel detector;(b) within the flat panel detector, performing the steps of,(1) generating digital data representative of digital images based on the electromagnetic radiation of the beam that is received by the flat panel detector, and(2) transmitting the digital data representative of digital images from the flat panel detector; and(c) at a computer, performing the steps of,(1) receiving the digital data representative of digital images transmitted from the flat panel detector, and(2) processing the digital data representative of digital images transmitted from the flat panel detector for display of digital images in the form of real-time video of the dental procedure;(d) using the processed data of said step (c)(2), displaying the real-time video of the dental procedure to a person at the location of the digital procedure;(e) wherein during said steps (a) through (d) the emitter and the flat panel detector do not concurrently rotate about the area in which the dental procedure is being performed;(f) wherein said steps (a) through (d) are performed such that the real-time video has a video frame rate of between 1 and 100 images per second; and(g) wherein the emitter is operated at voltage peaks within the range from 35 to 95 kVp and at current peaks between 0.0001 to 10 mA, and provides a beam with a continuous rate from 1 to 50 ms or with a pulse width range from 1 to 100 pulses/second, whereby low dose electromagnetic radiation is emitted during performance of said step (a). 2. The fluoroscopic imaging method of claim 1, wherein said step (d) comprises displaying the real-time video of the dental procedure to the person performing the procedure. 3. The fluoroscopic imaging method of claim 1, when the flat panel detector is an intraoral detector. 4. The fluoroscopic imaging method of claim 1, when the flat panel detector is an extraoral detector. 5. The fluoroscopic imaging method of claim 1, wherein the step performed at the computer of processing the data representative of digital images comprises transforming the digital data representative of digital images transmitted from the flat panel detector. 6. The fluoroscopic imaging method of claim 1, wherein the step performed at the computer of processing the data representative of digital images comprises recording the digital data representative of digital images transmitted from the flat panel detector. 7. The fluoroscopic imaging method of claim 1, wherein the step performed at the computer of processing the data representative of digital images comprises processing the digital data representative of digital images transmitted from the flat panel detector so as to enhance the represented digital images. 8. The fluoroscopic imaging method of claim 1, wherein the step performed at the computer of processing the data representative of digital images comprises compiling a series of the represented digital images into a video having a video frame rate ranging from 1 to 100 frames per second. 9. A fluoroscopic imaging method, comprising the steps of:(a) causing a beam of electromagnetic radiation to travel,(1) from a emitter,(2) through an area in which the dental procedure is being performed on a patient,(3) to a flat panel detector;(b) within the flat panel detector, performing the steps of,(1) transforming electromagnetic radiation of the beam that is received into electrical signals,(2) amplifying the electrical signals,(3) converting the amplified electrical signals into digital data representative of a digital image, and(4) transmitting the digital data representative of the digital image from the flat panel detector;(c) at a computer, performing the steps of,(1) receiving the digital data representative of the digital image transmitted from the flat panel detector, and(2) generating an image based at least in part on the received data representative of the digital image for display of digital images in the form of real-time video of the dental procedure; and(d) repeating said steps (a) through (c) and, using the generated images of said step (c)(2), displaying real-time video of the dental procedure to a person at the location of the digital procedure;(e) wherein during performance of said steps (a)-(d) the emitter and the flat panel detector do not concurrently rotate about the area in which the dental procedure is being performed; and(f) wherein during performance of said steps (a)-(d) the emitter is operated at voltage peaks within the range from 35 to 95 kVp and at current peaks between 0.0001 to 10 mA, and provides an beam with a continuous rate from 1 to 50 ms or with a pulse width range from 1 to 100 pulses/second, whereby low dose electromagnetic radiation is emitted during performance of said step (a). 10. The fluoroscopic imaging method of claim 9, further comprising, while performing said steps (a) through (c), repeatedly performing the steps of,(g) causing a second beam of electromagnetic radiation to travel,(1) from a second emitter,(2) through the area in which the dental procedure is being performed on the patient,(3) to a second flat panel detector;(h) within the second flat panel detector, performing the steps of,(1) transforming electromagnetic radiation of the second beam that is received into electrical signals,(2) amplifying the electrical signals,(3) converting the amplified electrical signals into digital data representative of a second digital image, and(4) transmitting the digital data representative of the second digital image from the second flat panel detector; and(i) at the computer, performing the steps of,(1) receiving the digital data representative of the second digital image transmitted from the second flat panel detector, and(2) generating an image based at least in part on the received data representative of the second digital image for the display of digital images in the form of the real-time video of the dental procedure;(j) wherein said step (d) further comprises using the generated images of said step (i)(2). 11. The fluoroscopic imaging method of claim 10, wherein said steps (a)-(i) are performed such that the real-time video has a video frame rate of between 1 and 100 images per second. 12. The fluoroscopic imaging method of claim 10, wherein said step (d) comprises displaying the real-time video of the dental procedure to the person performing the procedure. 13. The fluoroscopic imaging method of claim 10, wherein at least one of the two flat panel detectors is an intraoral detector. 14. The fluoroscopic imaging method of claim 10, wherein at least one of the two flat panel detectors is an extraoral detector. 15. The fluoroscopic imaging method of claim 10, wherein during performance of said steps the emitters and the flat panel detectors do not concurrently rotate about the area in which the dental procedure is being performed. 16. The fluoroscopic imaging method of claim 10, wherein during performance of said steps each of the emitters is operated at voltage peaks within the range from 35 to 95 kVp and at current peaks between 0.0001 to 10 mA, and provides an beam with a continuous rate from 1 to 50 ms or with a pulse width range from 1 to 100 pulses/second, whereby low dose electromagnetic radiation is emitted. 17. A fluoroscopic imaging method, comprising the steps of:(a) causing,(1) a first beam of electromagnetic radiation to travel,(A) from a first emitter,(B) through an area in which the dental procedure is being performed on a patient,(C) to a first flat panel detector;(2) a second beam of electromagnetic radiation to travel,(A) from a second emitter,(B) through the area in which the dental procedure is being performed on a patient,(C) to a second flat panel detector;(b) within the first flat panel detector, performing the steps of,(1) transforming electromagnetic radiation of the first beam that is received into electrical signals,(2) amplifying the electrical signals,(3) converting the amplified electrical signals into digital data representative of a first digital image, and(4) transmitting the digital data representative of the first digital image from the first flat panel detector;(c) within the second flat panel detector, performing the steps of,(1) transforming electromagnetic radiation of the second beam that is received into electrical signals,(2) amplifying the electrical signals,(3) converting the amplified electrical signals into digital data representative of a second digital image, and(4) transmitting the digital data representative of the second digital image from the second flat panel detector;(d) at a computer, performing the steps of,(1) receiving the digital data representative of the first digital image transmitted from the first flat panel detector and the digital data representative of the second digital image transmitted from the second flat panel detector, and(2) generating an image based at least in part on the received data representative of the first digital image and the received data representative of the second digital image for display of digital images in the form of real-time video of the dental procedure; and(e) repeating said steps (a) through (d) and, using the generated images of said step (d)(2), displaying real-time video of the dental procedure to a person at the location of the digital procedure; and,(f) wherein during performance of steps (a) through (d), the emitters and the flat panel detectors do not concurrently rotate about the area in which the dental procedure is being performed. 18. The fluoroscopic imaging method of claim 17, wherein the real-time video that is displayed is a 3D video. 19. The fluoroscopic imaging method of claim 17, wherein during performance of said steps each of the emitters is operated at voltage peaks within the range from 35 to 95 kVp and at current peaks between 0.0001 to 10 mA, and provides an beam with a continuous rate from 1 to 50 ms or with a pulse width range from 1 to 100 pulses/second, whereby low dose electromagnetic radiation is emitted. 20. A fluoroscopic imaging method, comprising the steps of:(a) causing a first beam of electromagnetic radiation to travel from a first emitter through an examination area to a first flat panel detector;(b) within the first flat panel detector, performing the steps of,(i) transforming electromagnetic radiation of the first beam that is received into electrical signals,(ii) amplifying the electrical signals,(iii) converting the amplified electrical signals into digital data representative of a first digital image, and(iv) transmitting the digital data representative of the first digital image from the first flat panel detector; and(c) at a computer, performing the steps of,(i) receiving the digital data representative of the first digital image transmitted from the first flat panel detector, and(ii) generating an image based at least in part on the received data representative of the first digital image; andfurther comprising the steps of,(d) causing a second beam of electromagnetic radiation to travel from a second emitter through the examination area to a second flat panel detector;(e) within the second flat panel detector, performing the steps of,(i) transforming electromagnetic radiation of the second beam that is received into electrical signals,(ii) amplifying the electrical signals,(iii) converting the amplified electrical signals into digital data representative of a second digital image, and(iv) transmitting the digital data representative of the second digital image from the second flat panel detector; and(f) at the computer, performing the steps of,(i) receiving the digital data representative of the second digital image transmitted from the second flat panel detector, and(ii) generating the image based at least in part on the received data representative of the second digital image;(g) wherein the first beam and the second beam intersect generally orthogonally at the examination area. 21. A fluoroscopic imaging method, comprising the steps of:(a) causing a first beam of electromagnetic radiation to travel from a first emitter through an examination area to a first flat panel detector;(b) within the first flat panel detector, performing the steps of,(i) transforming electromagnetic radiation of the first beam that is received into electrical signals,(ii) amplifying the electrical signals,(iii) converting the amplified electrical signals into digital data representative of a first digital image, and(iv) transmitting the digital data representative of the first digital image from the first flat panel detector; and(c) at a computer, performing the steps of,(i) receiving the digital data representative of the first digital image transmitted from the first flat panel detector, and(ii) generating an image based at least in part on the received data representative of the first digital image; andfurther comprising the steps of,(d) causing a second beam of electromagnetic radiation to travel from a second emitter through the examination area to a second flat panel detector;(e) within the second flat panel detector, performing the steps of,(i) transforming electromagnetic radiation of the second beam that is received into electrical signals,(ii) amplifying the electrical signals,(iii) converting the amplified electrical signals into digital data representative of a second digital image, and(iv) transmitting the digital data representative of the second digital image from the second flat panel detector; and(f) at the computer, performing the steps of,(i) receiving the digital data representative of the second digital image transmitted from the second flat panel detector, and(ii) generating the image based at least in part on the received data representative of the second digital image;(g) wherein the digital data representative of the first digital image is transmitted from the first flat panel detector at the same time that the digital data representative of the second digital image is transmitted from the second flat panel detector. 22. A fluoroscopic imaging method, comprising the steps of:(a) causing a first beam of electromagnetic radiation to travel from a first emitter through an examination area to a first flat panel detector;(b) within the first flat panel detector, performing the steps of,(i) transforming electromagnetic radiation of the first beam that is received into electrical signals,(ii) amplifying the electrical signals,(iii) converting the amplified electrical signals into digital data representative of a first digital image, and(iv) transmitting the digital data representative of the first digital image from the first flat panel detector; and(c) at a computer, performing the steps of,(i) receiving the digital data representative of the first digital image transmitted from the first flat panel detector, and(ii) generating an image based at least in part on the received data representative of the first digital image; andfurther comprising the steps of,(d) causing a second beam of electromagnetic radiation to travel from a second emitter through the examination area to a second flat panel detector;(e) within the second flat panel detector, performing the steps of,(i) transforming electromagnetic radiation of the second beam that is received into electrical signals,(ii) amplifying the electrical signals,(iii) converting the amplified electrical signals into digital data representative of a second digital image, and(iv) transmitting the digital data representative of the second digital image from the second flat panel detector; and(f) at the computer, performing the steps of,(i) receiving the digital data representative of the second digital image transmitted from the second flat panel detector, and(ii) generating the image based at least in part on the received data representative of the second digital image;(g) wherein the image is displayed to a user in real time. 23. A fluoroscopic imaging method, comprising the steps of:(a) causing a first beam of electromagnetic radiation to travel from a first emitter through an examination area to a first flat panel detector;(b) within the first flat panel detector, performing the steps of,(i) transforming electromagnetic radiation of the first beam that is received into electrical signals,(ii) amplifying the electrical signals,(iii) converting the amplified electrical signals into digital data representative of digital images, and(iv) transmitting the digital data representative of digital images from the first flat panel detector; and(c) at a computer, performing the steps of,(i) receiving the digital data representative of digital images transmitted from the first flat panel detector, and(ii) displaying a video to a user in real based at least in part on the digital data representative of digital images that is received from the first flat panel detector; andfurther comprising the steps of,(d) causing a second beam of electromagnetic radiation to travel from a second emitter through the examination area to a second flat panel detector;(e) within the second flat panel detector, performing the steps of,(i) transforming electromagnetic radiation of the second beam that is received into electrical signals,(ii) amplifying the electrical signals,(iii) converting the amplified electrical signals into digital data representative of a second digital image, and(iv) transmitting the digital data representative of the second digital image from the second flat panel detector; and(f) at the computer, performing the steps of,(i) receiving the digital data representative of digital images transmitted from the second flat panel detector, and(ii) displaying the video to the user in real based in part on the digital data representative of digital images that is received from the second flat panel detector.
	description	The present invention relates, in general, to intelligent assets, and more particularly, to maintaining intelligent assets. Assets are devices that an enterprise uses to perform business operations, e.g., various processing, assembly, transportation and storage devices on various production lines. Intelligent assets are devices having built-in computing and communication capabilities, and they are capable of sending data, such as their own locations, status, performance, etc., to an asset management system. An asset management system collects and stores data from assets for status checking, maintenance, problem diagnosis and repair of the assets. Asset management systems directly collect the data of the locations, statuses, performances, etc. of the assets, for example, the assets' temperatures, power consumptions and running speeds, etc., from the assets. It can be determined whether the assets are running or malfunctioning, but a degree in the reduction of the work efficiency cannot be determined, and, thus, a more timely and accurate asset maintenance plan cannot be formed. According to one embodiment of the present invention, a system for maintaining intelligent assets comprises a processor and memory connected to the processor. A work performance acquisition module obtains work performance data of the intelligent assets from a device operation production control system. A status monitoring module obtains operation status data of the intelligent assets. A loss calculation module calculates a loss degree of the intelligent assets according to the obtained work performance and operation status data, and a maintenance determining module determines whether the intelligent assets need maintenance according to the loss degree of the intelligent assets. According to one embodiment of the present invention, a method for maintaining intelligent assets obtains work performance data of the intelligent assets from a device operation production control system. Operation status data of the intelligent assets is obtained. A processor calculates a loss degree of the intelligent assets according to the obtained work performance data and the operation status data. A determination is made as to whether the intelligent assets need maintenance according to the calculated loss degree of the intelligent assets. According to one embodiment of the present invention, a computer program product for maintaining intelligent assets comprises a computer readable storage medium having computer readable program code embodied therewith. The computer readable program code comprises computer readable program code configured to obtain work performance data of the intelligent assets from a device operation production control system. Computer readable program code is configured to obtain operation status data of the intelligent assets. Computer readable program code is configured to calculate a loss degree of the intelligent assets according to the obtained work performance data and the operation status data. Computer readable program code is configured to determine whether the intelligent assets need maintenance according to the calculated loss degree of the intelligent assets. As will be appreciated by one skilled in the art, aspects of the present invention may be embodied as a system, method or computer program product. Accordingly, aspects of the present invention may take the form of an entirely hardware embodiment, an entirely software embodiment (including firmware, resident software, micro-code, etc.) or an embodiment combining software and hardware aspects that may all generally be referred to herein as a “circuit,” “module” or “system.” Furthermore, aspects of the present invention may take the form of a computer program product embodied in one or more computer readable medium(s) having computer readable program code embodied thereon. Any combination of one or more computer readable medium(s) may be utilized. The computer readable medium may be a computer readable signal medium or a computer readable storage medium. A computer readable storage medium may be, for example, but not limited to, an electronic, magnetic, optical, electromagnetic, infrared, or semiconductor system, apparatus, or device, or any suitable combination of the foregoing. More specific examples (a non-exhaustive list) of the computer readable storage medium would include the following: an electrical connection having one or more wires, a portable computer diskette, a hard disk, a random access memory (RAM), a read-only memory (ROM), an erasable programmable read-only memory (EPROM or Flash memory), an optical fiber, a portable compact disc read-only memory (CD-ROM), an optical storage device, a magnetic storage device, or any suitable combination of the foregoing. In the context of this document, a computer readable storage medium may be any tangible medium that can contain, or store a program for use by or in connection with an instruction execution system, apparatus, or device. A computer readable signal medium may include a propagated data signal with computer readable program code embodied therein, for example, in baseband or as part of a carrier wave. Such a propagated signal may take any of a variety of forms, including, but not limited to, electro-magnetic, optical, or any suitable combination thereof. A computer readable signal medium may be any computer readable medium that is not a computer readable storage medium and that can communicate, propagate, or transport a program for use by or in connection with an instruction execution system, apparatus, or device. Program code embodied on a computer readable medium may be transmitted using any appropriate medium, including but not limited to wireless, wireline, optical fiber cable, RF, etc., or any suitable combination of the foregoing. Computer program code for carrying out operations for aspects of the present invention may be written in any combination of one or more programming languages, including an object oriented programming language such as Java, Smalltalk, C++ or the like and conventional procedural programming languages, such as the “C” programming language or similar programming languages. The program code may execute entirely on the user's computer, partly on the user's computer, as a stand-alone software package, partly on the user's computer and partly on a remote computer or entirely on the remote computer or server. In the latter scenario, the remote computer may be connected to the user's computer through any type of network, including a local area network (LAN) or a wide area network (WAN), or the connection may be made to an external computer (for example, through the Internet using an Internet Service Provider). Aspects of the of the present invention are described below with reference to flowchart illustrations and/or block diagrams of methods, apparatus (systems) and computer program products according to embodiments of the invention. It will be understood that each block of the flowchart illustrations and/or block diagrams, and combinations of blocks in the flowchart illustrations and/or block diagrams, can be implemented by computer program instructions. These computer program instructions may be provided to a processor of a general purpose computer, special purpose computer, or other programmable data processing apparatus to produce a machine, such that the instructions, which execute via the processor of the computer or other programmable data processing apparatus, create means for implementing the functions/acts specified in the flowchart and/or block diagram block or blocks. These computer program instructions may also be stored in a computer readable medium that can direct a computer, other programmable data processing apparatus, or other devices to function in a particular manner, such that the instructions stored in the computer readable medium produce an article of manufacture including instructions which implement the function/act specified in the flowchart and/or block diagram block or blocks. The computer program instructions may also be loaded onto a computer, other programmable data processing apparatus, or other devices to cause a series of operational steps to be performed on the computer, other programmable apparatus or other devices to produce a computer implemented process such that the instructions which execute on the computer or other programmable apparatus provide processes for implementing the functions/acts specified in the flowchart and/or block diagram block or blocks. The flowchart and block diagrams in the Figures illustrate the architecture, functionality, and operation of possible implementations of systems, methods and computer program products according to various embodiments of the present invention. In this regard, each block in the flowchart or block diagrams may represent a module, segment, or portion of code, which comprises one or more executable instructions for implementing the specified logical function(s). It should also be noted that, in some alternative implementations, the functions noted in the block may occur out of the order noted in the figures. For example, two blocks shown in succession may, in fact, be executed substantially concurrently, or the blocks may sometimes be executed in the reverse order, depending upon the functionality involved. It will also be noted that each block of the block diagrams and/or flowchart illustration, and combinations of blocks in the block diagrams and/or flowchart illustration, can be implemented by special purpose hardware-based systems that perform the specified functions or acts, or combinations of special purpose hardware and computer instructions. The terminology used herein is for the purpose of describing particular embodiments only and is not intended to be limiting of the invention. As used herein, the singular forms “a”, “an” and “the” are intended to include the plural forms as well, unless the context clearly indicates otherwise. It will be further understood that the terms “comprises” and/or “comprising,” when used in this specification, specify the presence of stated features, integers, steps, operations, elements, and/or components, but do not preclude the presence or addition of one or more other features, integers, steps, operations, elements, components, and/or groups thereof. FIG. 1 is one example of a system 10 for maintaining intelligent assets according to an embodiment of the present invention. As shown, the system 10 comprises the following modules: a work performance data acquiring module 11 configured to obtain work performance data of intelligent assets from a device operation production control system 20; a status monitoring module 12 for obtaining operation status data of the intelligent assets; a loss calculation module 13 configured to calculate a loss degree of the intelligent assets according to the obtained work performance and operation status data; and a maintenance determining module 14 configured to determine whether the intelligent assets need maintenance according to the loss degree of the intelligent assets. The system 10 further includes a processor 16. Intelligent assets can be any set of intelligent assets for accomplishing specific business objectives, e.g., they can be work stations in a production line, water pumps in a water treatment process, conveying equipment, etc. The device operation production control system 20 can be any controller for controlling the operation of the intelligent assets to accomplish their business objectives, e.g., an assembly line controller, a pump controller, etc. The system 10 for maintaining intelligent assets according to an embodiment of the present invention can obtain work performance data from the intelligent assets from the device operation production control system 20 via a network connection. The device operation production control system 20 may further include a processor 17. The work performance data may be the workload or output accomplished by the intelligent assets, e.g., the number of work pieces processed by a work station, or the discharge of a water pump. The system 10 can obtain the operation status data of the intelligent assets from the intelligent assets directly via the network connection with the intelligent assets, or, of course, can obtain the operation status data of the intelligent asset via other devices, e.g., the device operation production control system 20. Various status data, such as power consumption, running speed and temperature, etc., of the intelligent assets can be obtained from the intelligent assets by using various sensors. The device operation production control system 20 is connected via a network with the intelligent assets, controls the operation of the intelligent assets to accomplish the business objectives, and collects the work performance data from the intelligent assets. According to an embodiment of the present invention, work performance data refers to the workload accomplished by the intelligent assets, and the operation status data is the power consumption of the intelligent assets. The loss calculation module 13 is configured to calculate a loss degree of the intelligent assets according to the obtained work performance data, and the operation status data is further configured to: calculate the ratio between the workload accomplished by the intelligent assets and the power consumption; calculate a loss degree of the intelligent assets according to the calculated ratio and the ratio between the workload accomplished by the intelligent assets at an initial time and the power consumption. The initial time refers to the time when an intelligent asset is just put into operation and, thus, in excellent condition. At this time, the work efficiency of the intelligent asset is the highest, and, thus, the ratio between the workload accomplished by the intelligent asset (e.g., number of work pieces processed by the work stations, water amount pumped by the water pump, etc.) and the power consumption is the largest. With wear of the intelligent assets, its work efficiency will be gradually reduced, and the ratio between its accomplished workload and the power consumption is reduced correspondingly. Therefore, the loss degree of the intelligent assets can be determined according to the degree of reduction of the ratio between the workload accomplished by the intelligent assets and the power consumption as compared with the ratio at the initial time. When the calculated ratio is smaller than the ratio at the initial time, it can be determined that the intelligent asset has a loss in efficiency. The bigger the extent to which the calculated ratio is smaller than the initial ratio is, the bigger the loss degree of the intelligent asset is. When the calculated ratio is smaller than the ratio of the initial time by up to a predetermined threshold, e.g., to 20% or 30%, it can be determined that the intelligent asset has a severe loss, and thus needs to be repaired or replaced. The loss of an intelligent asset will not only cause a reduction of work efficiency, but it will also cause a reduction of work performance of the intelligent asset under a specific work condition. Therefore, the loss degree of the intelligent asset can be determined by comparing the work performance of the intelligent asset in the current specific work condition and the work performance of the intelligent asset in the initial specific condition. The specific work condition can be, e.g., the highest temperature, or a state of full open of the pump. If the work performance of the intelligent asset under the current specific work condition is lower than that under the initial specific work condition, it can be determined that the intelligent asset has a loss; the more the work performance of the intelligent asset under the current specific work condition is reduced as compared with the work performance of the intelligent asset under the initial specific work condition, the bigger the loss degree of the intelligent asset is; if the work performance is reduced by up to a predetermined threshold, it can be determined that the intelligent asset has a severe loss, and needs to be repaired or replaced. Therefore, according to another embodiment of the present invention, the work performance data is the workload accomplished by the intelligent asset, and the operation status data represents the specific work condition of the intelligent asset, and the calculation module 13 for calculating the loss degree of the intelligent asset according to the obtained work performance data and the operation status data is further configured to: calculate a loss degree of the intelligent asset by comparing the workload accomplished by the intelligent asset currently under the specific work condition with the workload accomplished by the intelligent asset initially under the specific work condition. The system 10 for maintaining intelligent assets according to an embodiment of the present invention may further consider a group of intelligent assets (e.g., a group of intelligent assets which form a production line) as a larger intelligent asset, and calculates a loss degree of the larger intelligent asset according to the status data and work performance of the larger intelligent asset. For example, the work performance data of the entire production line may be the output of the production line, the status data of the production line may be the sum of the status data of all the assets in the production line, e.g., a sum of the power consumption of all the intelligent assets. The system 10 for maintaining intelligent assets according to an embodiment of the present invention may calculate the ratio between the output of the entire production line and the power consumption of the entire production line, and compare the ratio with the ratio between the output and the power consumption of the production line at the initial time. When the calculated ratio is smaller than the initial ratio, it can be determined that the production line has a loss. The bigger the extent to which the calculated ratio is smaller than the initial ratio is, the bigger the loss of the production line can be determined. When the calculated ratio is smaller than the ratio in the normal condition by up to a predetermined threshold, e.g., 20% or 30%, it can be determined that the production line has a severe loss, and thus its component assets need to be repaired or replaced. At this time, each intelligent asset, which is determined as having a severe or relatively severe loss, can be repaired or replaced one by one, and after each repair or replacement, the same method can be used to determine whether the loss of the entire production line is restored to be smaller than the threshold. When it is determined that the loss of the entire production line is smaller than the threshold, the repair or replacement of the remaining intelligent assets can be cancelled. According to an embodiment of the present invention, the maintenance determining module 14 determines that an intelligent asset needs to be maintained in response to determining that the loss degree of the intelligent asset is larger than a predetermined threshold. According to a further embodiment of the present invention, the system 10 for maintaining intelligent assets further comprises an optional notification module 15 configured to, in response to determining that the intelligent asset needs maintenance, send a notification for maintaining the intelligent asset to the device operation production control system 20. According to an embodiment of the present invention, the device operation production control system 20 may maintain the intelligent asset according to the received notification for maintaining the intelligent asset. For example, the device operation production control system 20 may set the status of the corresponding intelligent asset as “to be maintained” according to the received notification, so as to stop the operation of the intelligent asset. According to an embodiment of the present invention, the device operation production control system 20 may maintain the intelligent asset according to the notification and the relationship between the intelligent asset and other intelligent assets. For example, for an intelligent asset having backup assets, the backup assets may be initiated first to replace the intelligent asset, and then the status of the intelligent asset may be set as “to be maintained,” thus, the operation of the production line will not be interrupted. For an intelligent asset without backup assets, the status of the intelligent asset is set as “to be maintained,” and the operation of the entire production line has to be stopped. According to some other embodiments of the present invention, the system 10 for maintaining intelligent assets may send the notification for maintaining the intelligent assets to other external or internal devices, so that the other external or internal devices may perform maintenance on the intelligent assets. According to some embodiments of the present invention, the system 10 for maintaining intelligent assets resides in the asset management system. According to some other embodiments of the present invention, the system 10 for maintaining intelligent assets resides outside the asset management system. According to some yet other embodiments of the present invention, the system 10 for maintaining intelligent assets resides in both the asset management system and the device operation production control system. Above is described the structure of the system for maintaining intelligent assets according to embodiments of the present invention by referring to the accompanying drawings. It should be pointed out that the above description and illustration are merely exemplary and illustrative, and not limitation to the present invention. The asset management system and the device operation production control system, according to the embodiments of the present invention may have more, less or different modules, and the connection and containment relationship among the modules may be different from what is illustrated and described. A method for maintaining intelligent assets according to an embodiment of the present invention is now discussed by referring to FIG. 2. As shown, the method comprises the following steps: In step 21, work performance data of intelligent assets is obtained from a device operation production control system. In step 22, operation status data of the intelligent assets is obtained. In step 23, a loss degree of the intelligent assets is calculated according to the obtained work performance data and operation status data. In step 24, it is determined whether the intelligent assets need maintenance according to the calculated loss degree. According to an embodiment of the present invention, step 24 for determining whether the intelligent assets needs maintenance according to the calculated loss degree comprises the following sub-step: in response to determining that the loss degree of the intelligent assets is larger than a predetermined threshold, determining that the intelligent assets need maintenance. According to an embodiment of the present invention, the method further comprises the following step: in response to determining that the intelligent assets need maintenance, sending a notification for maintaining the intelligent assets to the device operation production control system. According to an embodiment of the present invention, the work performance data is the workload accomplished by the intelligent assets, and the operation status data is the power consumption of the intelligent assets, and step 23 for calculating the loss degree of the intelligent assets according to the obtained work performance data and the operation status data comprises the following sub-steps: calculating the ratio between the workload accomplished by the intelligent assets and the power consumption; calculating a loss degree of the intelligent assets according to the calculated ratio and the ratio between the workload accomplished by the intelligent assets and the power consumption at an initial time. According to an embodiment of the present invention, the work performance data is the workload accomplished by the intelligent assets, and the operation status data represents a specific work condition of the intelligent assets, and step 23 for calculating the loss degree of the intelligent assets according to the obtained work performance data and the operation status data comprises: calculating a loss degree of the intelligent assets by comparing the workload accomplished by the intelligent assets currently under the specific work condition with the workload accomplished by the intelligent assets initially under the specific work condition. According to an embodiment of the present invention, the method further comprises a step that the device operation production control system maintains the intelligent assets according to the received notification of maintaining the intelligent assets. According to an embodiment of the present invention, the step that the device operation production control system maintains the intelligent assets according to the notification comprises: maintaining, by the device operation production control system, the intelligent assets according to the notification and relationships between the intelligent assets and other intelligent assets. According to some embodiments of the present invention, the method is executed by the asset management system. According to some other embodiments of the present invention, the method is executed by other devices outside the asset management system. According to some yet other embodiments, the method is executed by the asset management system and the device operation production control system. Above is described a method for maintaining intelligent assets according to embodiments of the present invention. It should be pointed out that the above description and illustration are only exemplary and illustrative, rather than as limitations to the present invention. The method for maintaining intelligent assets according to embodiments of the present invention may have more, fewer or different steps, and the relationships between the steps can be different from what is illustrated or described. The corresponding structures, materials, acts, and equivalents of all elements in the claims below are intended to include any structure, material, or act for performing the function in combination with other claimed elements as specifically claimed. The description of the present invention has been presented for purposes of illustration and description, but is not intended to be exhaustive or limited to the invention in the form disclosed. Many modifications and variations will be apparent to those of ordinary skill in the art without departing from the scope and spirit of the invention. The embodiment was chosen and described in order to best explain the principles of the invention and the practical application, and to enable others of ordinary skill in the art to understand the invention for various embodiments with various modifications as are suited to the particular use contemplated. Having thus described the invention of the present application in detail and by reference to embodiments thereof, it will be apparent that modifications and variations are possible without departing from the scope of the invention defined in the appended claims.
052326586	claims	1. A fuel assembly for a boiling water reactor, comprising an elongated box having box walls with flat outer surfaces defining a polygonal outer cross section with rounded corners, an interior and a longitudinal axis, a bundle of fuel rods being disposed in said box and aligned parallel to one another and to the longitudinal axis of said box, said fuel rods being disposed beside one another in rows parallel to said box walls, said box walls having reinforcements protruding into said interior in the vicinity of said rounded corners defining a relatively increased wall thickness, and said box walls having a relatively reduced wall thickness between said reinforcements, said bundle of fuel rods having no fuel rod at intersections of two of said rows of fuel rods adjacent to two of said box walls bordering one another at said rounded corners. 2. The fuel assembly according to claim 1, wherein said elongated box extends between a top part and a bottom part. 3. The fuel assembly according to claim 1, wherein two of said fuel rods in said rows within said bundle are spaced between apart by a minimum spacing being equal to a spacing between two fuel rods in a row parallel to one of said box walls. 4. The fuel assembly according to claim 1, wherein said fuel rods of two adjacent rows are staggered relative to one another in the direction of said box walls. 5. The fuel assembly according to claim 1, wherein said relatively reduced wall thickness is less than approximately 2.4 mm, and said relatively increased wall thickness in the vicinity of said rounded corners is between approximately 2.5 and 3.0 mm. 6. The fuel assembly according to claim 1, wherein said relatively reduced wall thickness is less than approximately between 1.5 and 1.7 mm, and said relatively increased wall thickness in the vicinity of said rounded corners is between approximately 2.7 to 2.9 mm. 7. In a boiling water reactor having a core zone with a given geometry and a plurality of fuel assemblies, each of said fuel assemblies comprising an elongated box having box walls with flat outer surfaces defining a polygonal outer cross section with rounded corners, an interior and a longitudinal axis, said outer cross section being defined by the given geometry of the core zone, a bundle of fuel rods being disposed in said box and aligned parallel to one another and to the longitudinal axis of said box, said fuel rods being disposed beside one another in rows parallel to said box walls, said box walls having reinforcements protruding into said interior in the vicinity of said rounded corners defining a relatively increased wall thickness, and said box walls having a relatively reduced wall thickness between said reinforcements, said bundle of fuel rods having no fuel rod at intersections of two of said rows of fuel rods adjacent to two of said box walls bordering one another at said rounded corners.
	abstract	A canister for storing radioactive materials includes a base plate, side wall and a top plate. The top plate includes a top surface with a top edge having a bevel, and with a channel set in from the top edge. The top plate is sealed to the sidewall by a weld formed between the beveled top edge and the top of the side wall. The base plate is sealed to a bottom of the sidewall, so that a sealed vessel is formed.
	claims	1. A collimator for a radiation detector including at least three spacing elements arranged on a radiation exit face of the collimator; wherein the collimator, for scattered radiation reduction in a φ- and in z-direction, is embodied as a 2D collimator with a cell-type structure; and wherein the collimator is manufactured by way of a selective laser melting or by way of an injection molding method with plastic, which is manufactured with an x-ray absorbing powder. 2. The collimator as claimed in claim 1, wherein the spacing elements in the direction of radiation incidence are dimensioned according to a total of a maximum number of expected manufacturing-related deviations of the radiation exit face and of a support surface of a radiation converter of the radiation detector from planned surfaces. 3. The collimator as claimed in claim 2, wherein the collimator with the spacing elements is a one-piece element. 4. The collimator as claimed in claim 2, wherein the spacing elements are arranged in corner positions or on positions of edge bisectors of the radiation exit face of the collimator. 5. The collimator as claimed in claim 2, wherein at least one spacing element is formed by at least one absorber element projecting out of the radiation exit face of the collimator. 6. The collimator as claimed in claim 1, wherein the collimator with the spacing elements is a one-piece element. 7. The collimator as claimed in claim 1, wherein the collimator is manufactured by way of a Rapid Manufacturing technique. 8. The collimator as claimed in claim 1, wherein the spacing elements are arranged in corner positions or on positions of edge bisectors of the radiation exit face of the collimator. 9. The collimator as claimed in claim 1, wherein at least one spacing element is formed by at least one absorber element projecting out of the radiation exit face of the collimator. 10. A method for manufacturing a collimator for a radiation detector, the method comprising:embodying absorber elements, crossing over each other, layer by layer from a radiation-absorbing material along a φ- and z-direction by way of a Rapid Manufacturing technique or by way of an injection molding method using plastic, which is filled with an x-ray absorbing powder, and wherein at least three spacing elements are embodied in addition on a radiation exit face of the collimator. 11. The method as claimed in claim 10, wherein the spacing elements are dimensioned according to a total of the maximum number of expected manufacturing-related deviations of the radiation exit face and of a mounting surface of a radiation converter of the radiation detector from planned surfaces in the direction of radiation incidence.
060884193	description	DETAILED DESCRIPTION OF THE INVENTION With the foregoing and other objectives in view there is provided, in accordance with the invention, a nuclear fuel rod comprising a cladding tube of a zirconium alloy and having a highly corrosion resistant outer portion in which zirconium hydride precipitation is inhibited and an inner portion in which zirconium hydride precipitation is promoted. In accordance with the present invention, there is also provided a nuclear fuel rod comprising a multiple-layer composite cladding having an inner zirconium or zirconium alloy layer and a highly corrosion resistant zirconium or zirconium alloy outer layer having an outer portion in which hydride precipitation is inhibited and an inner portion of the outer layer in which zirconium hydride precipitation is promoted. In accordance with the present invention, there is further provided a nuclear fuel rod comprising a multi-layered composite cladding having a highly corrosion resistant zirconium or zirconium alloy outer layer in which hydride precipitation is inhibited, and a zirconium or zirconium alloy inner layer where hydride precipitation is promoted. In a recent research program involving corrosion studies of a composite two-layer cladding for a nuclear fuel rod for a light water reactor having an outer layer of Zircaloy 4 and a zirconium inner layer, it was observed by the present inventor that zirconium hydrides precipitated in the lower oxygen content zirconium inner layer and not in the outer higher oxygen content Zircaloy 4 layer whereas the temperature gradient across the wall of the composite cladding was such that hydride precipitation would have been expected in the Zircaloy 4 outer layer. The present inventor discovered that the amount of hydride precipitation is in part a function of the amount of oxygen in the zirconium alloy and zirconium metal and by limiting the oxygen content to a low level in an inner portion or layer of a zirconium metal and/or zirconium alloy fuel rod cladding and increasing the oxygen content to a higher level in an outer portion or layer of the cladding, enhanced resistance to hydride formation and corrosion is obtained in the outer portion or layer compared to conventional single layer or multiple layered claddings made of zirconium and/or zirconium alloys with a non-varying or uniform oxygen content across the cladding wall. During continued corrosion of the cladding when exposed in a nuclear reactor, the absence of hydride formation near the cladding outside surface limits the corrosion reaction to that described by Equation (1), whereas if hydrides precipitate in large quantities near the outside surface of the cladding, the corrosion reaction is more appropriately described by Equation (2) which, as explained above, leads to an acceleration in the corrosion rate. In accordance with the present invention, by inhibiting the formation of hydride precipitates in the outer portion or layer of the cladding, accelerated corrosion is inhibited. In accordance with the present invention, a nuclear fuel rod for water moderated or cooled reactors is provided having a metallic tubular cladding comprising a zirconium alloy and having a decreasing oxygen concentration gradient from the outer wall to the inner wall where despite being subject to an increasing temperature gradient from the outer wall to the inner wall of the cladding during reactor operations, zirconium hydrides preferentially precipitate in the inner portions of the cladding away from the cladding outside wall and are inhibited from forming on the outer wall. In a preferred embodiment, the oxygen content in the cladding tube decreases from an amount greater than approximately 1600 ppm at the outer wall of the cladding to less than approximately 1200 ppm at the inner wall of the cladding. In an alternative embodiment, a nuclear fuel rod for water moderated or cooled reactors is provided having a cladding tube comprising a composite of two or more layers of zirconium and/or zirconium alloy metals, the outermost layer having a higher oxygen content than an inner layer of the cladding where despite being subject to an increasing temperature gradient across the cladding wall during reactor operations, zirconium hydrides will preferentially precipitate in the inner layer of the cladding and will be preferentially inhibited from precipitating in the outermost layer of the cladding. In a preferred embodiment, the oxygen content of the outermost layer is greater than about 1600 ppm and the oxygen content of the inner layer is less than about 1200 ppm. In another alternative embodiment, a nuclear fuel rod for water moderated or cooled reactors is provided having a cladding tube comprising a composite of two or more layers, the outermost layer comprising a zirconium or zirconium alloy metal having a decreasing oxygen concentration gradient from the outer wall to the inner portion of the outermost layer, where despite a temperature gradient across the cladding wall, zirconium hydrides will preferentially precipitate in the inner portion of the outermost layer of the cladding and will be preferentially inhibited from precipitating near the outer wall of the outermost layer of the cladding. In a preferred embodiment, the oxygen content in the outermost layer of the composite cladding decreases from an amount greater than about 1600 ppm at the outer wall to less than about 1200 ppm in the inner portion of the outer layer of the composite cladding. In the outer portion or layer of the cladding comprised of the zirconium metal or alloy having the higher oxygen concentration, the hydrogen which was formed as a result of the reaction between the reactor coolant water and the zirconium and which was picked up in the outer portion or layer, continues to diffuse through the outer portion or layer and into the inner portion or layer of the cladding. The hydrogen does not precipitate to form hydrides in the oxygen-enriched outer portion or layer because the hydrogen concentration does not reach the raised solubility limit and diffuses into the low oxygen inner portion or layer where it preferentially precipitates as a result of the hydrogen concentration exceeding the lowered solubility limit. Opposing the preferential precipitation of hydrogen in the inner portion or layer low oxygen alloy is the temperature gradient across the wall of the cladding tube. Since the portions of the cladding closer to the fuel pellet are at a higher temperature than the portions of the cladding closer to or in contact with the reactor coolant water, the inner portion or layer low oxygen concentration alloy is at a higher temperature which raises the solubility limit of hydrogen in the metal alloy. Thus, the temperature gradient across the wall of the cladding tube affects the solubility limits of the inner layer and the outer layer and tends to encourage the precipitation of hydrogen as hydrides in the inner layer low oxygen metal alloy near the interface with the outer layer high oxygen metal alloy. Therefore, even in the presence of a temperature gradient which would promote hydride precipitation in the outermost layer of conventional cladding, such precipitation near the outside surface is prevented when the outer layer of the cladding comprises a high oxygen alloy and the inner portion of the cladding has a lower oxygen content. FIG. 1 represents a nuclear fuel assembly 10 for a pressurized water reactor (PWR) comprising a lower tie plate 12, guide tubes 14, nuclear fuel rods 18 which are spaced radially and supported by spacer grids 16 spaced along the guide tubes, an instrumentation tube 28, and an upper tie plate 26 attached to the upper ends of the guide tubes. Each fuel rod 18 generally includes a metallic tubular fuel rod cladding 100 within which are nuclear fuel pellets 80 composed of fissionable and/or fertile material and an upper end plug 22 and a lower end plug 24 which hermetically seal the nuclear fuel pellets within the metallic tubular fuel rod cladding as shown in FIG. 2. A helical spring member 21 can be positioned within the fuel rod to maintain the position of the fuel pellets in a stacked relationship. Control rods which are used to assist in controlling the fission reaction are disposed in the guide tubes, but are not shown. Referring to FIG. 3 which is a schematic representation of cross-sectional view of the nuclear fuel rod shown in FIG. 2, cladding 100 is a metallic tube having a single metal layer 101 of a zirconium alloy with a decreasing oxygen concentration gradient (depicted as a decreasing density in stippling) from outer wall 102 to inner wall 103. During reactor operations, fuel pellets 80 which are positioned within the cladding generate heat which is transferred through the cladding to outer wall 102 to the reactor coolant which results in a decreasing temperature gradient from inner wall 103 to outer wall 102 of the cladding. In a preferred embodiment, the oxygen concentration gradient from outer wall 102 to inner wall 103 decreases from at least about 1600 ppm to less than about 1200 ppm. Referring to FIG. 4A which is a schematic representation of a cross-sectional view of a nuclear fuel rod for a PWR such as shown in FIGS. 1 and 2 but instead has a composite cladding 110 comprising an outer layer 111 and an inner layer 114 each of which is composed of a zirconium and/or zirconium alloy metal. Outer layer 111 has a higher oxygen content than inner layer 114 (which is depicted as different stippling) so that hydrides will preferentially precipitate in inner layer 114 and be inhibited from precipitating in outer layer 111. In a preferred embodiment, outer layer 111 has an oxygen content greater than about 1600 ppm, and inner layer 25 has an oxygen content less than about 1200 ppm. Referring to FIG. 4B which is a schematic representation of a cross-sectional view of another nuclear fuel rod for a PWR such as shown in FIGS. 1 and 2 but instead has a composite cladding 120 comprising an outer layer 121, an inner layer 124 and an innermost layer 127. Outer layer 121 and inner layer 124 are composed of a zirconium and/or zirconium alloy metal. Outer layer 121 has a higher oxygen content than inner layer 124 (which is depicted as different stippling) so that hydrides will preferentially precipitate in inner layer 124 and be inhibited from precipitating in outer layer 121. Innermost layer 127 can be zirconium or a zirconium alloy, or another metal. In a preferred embodiment, outer layer 121 has an oxygen content greater than about 1600 ppm, and inner layer 124 has an oxygen content less than about 1200 ppm. Referring to FIG. 4C which is a schematic representation of a cross-sectional view of another nuclear fuel rod for a PWR such as shown in FIGS. 1 and 2 but instead has a composite cladding 130 comprising an outer layer 131, an inner layer 134 and an innermost layer 137 each of which is composed of a zirconium and/or zirconium alloy metal. Outer layer 131 has a higher oxygen content than inner layer 134 (which is depicted as different stippling) so that hydrides will preferentially precipitate in inner layer 134 and be inhibited from precipitating in outer layer 131. In a preferred embodiment, outer layer 131 has an oxygen content greater than about 1600 ppm, and inner layer 134 as an oxygen content less than about 1200 ppm. In another preferred embodiment, innermost layer 137 has an oxygen concentration which is at least that of inner layer 134 but less than or equal to outer layer 131. Referring to FIG. 5 which is a schematic representation of a cross-sectional view of another nuclear fuel rod for a PWR such as shown in FIGS. 1 and 2 but instead has a composite cladding 140 which comprises at least two layers of zirconium and/or zirconium alloy metals, including an outermost layer 141 and an inner layer 144. Outermost layer 141 has a decreasing oxygen concentration gradient depicted as a decreasing variation in stippling from outer wall 142 to an inner portion 141a, where despite a temperature gradient across the cladding wall, zirconium hydrides will preferentially precipitate in inner portion 141a of outermost layer 141 of cladding 140 and will be preferentially inhibited from precipitating near outer wall 142 of outermost layer 141 of cladding 140. In an alternative embodiment, composite cladding can include an innermost layer formed from a zirconium metal or alloy, or another metal or alloy. In a preferred embodiment, the innermost layer is a zirconium metal or alloy, and the oxygen content in outermost layer 141 of cladding 140 decreases from an amount greater than about 1600 ppm at outer wall 142 to less than about 1200 ppm in the inner portion 141a of outer layer 141. Referring to FIG. 6, a nuclear fuel assembly for a boiling water reactor (BWR) in the U.S. is generally shown at 30 having nuclear fuel rods 32 which are supported between a lower tie plate 34A and upper tie plate 36. Each fuel rod generally includes a metallic tubular fuel rod cladding 150 within which are nuclear fuel pellets 80 which are hermetically sealed within the tubular cladding by end sealing means such as end plugs. Lower tie plate 34A and upper tie plate 36 are connected structurally by tie rods 40 positioned within the array of fuel rods or by other means such as an inner water channel. Spacer grids 38 provide intermediate support of the fuel rods 32 over the length of the fuel assembly and maintain them in a spaced relationship while restraining them from lateral vibration. Outer channel 42 surrounds the fuel assembly and extends from the lower tie plate to the upper tie plate. An example of nuclear fuel assembly for use in boiling water reactors outside the U.S. and typically in Europe is generally shown at 30 in FIG. 7 and similarly has tie rods 40, spacer grids 38, outer channel 42, and fuel rods 32 each generally including a metallic tubular fuel cladding 150 within which are nuclear fuel pellets 80. The fuel rods 32 are supported between a lower tie plate 34B and upper tie plate 36. Referring to FIG. 8, nuclear fuel rod 32 shown in FIGS. 6 and 7 includes nuclear fuel shown as a plurality of fuel pellets 80 of fissionable and/or fertile material positioned within a metallic tubular fuel rod cladding 150. The metallic tubular fuel rod cladding is sealed at its ends by means of end plugs 54 which may include alignment pins 33 to facilitate the mounting of the fuel rod in the assembly. A void space or plenum 56 is provided at one end of the fuel rod to permit longitudinal expansion of the nuclear fuel and accumulation of gases released from the nuclear fuel. A helical spring member 58 is positioned within space 56 and is capable of maintaining the position of the fuel pellets during handling and transportation of the fuel rods. Cladding 150 is secured to end plugs 54 by means of circumferential welds 62. Referring to FIG. 9 which is a schematic representation of a cross-sectional view of the nuclear fuel rod shown in FIG. 8, cladding 150 is a metallic tube having a single metal layer 151 of a zirconium alloy with a decreasing oxygen concentration gradient (depicted as a decreasing density in stippling) from outer wall 152 to inner wall 153. In a preferred embodiment, the oxygen concentration gradient from outer wall 152 to inner wall 153 decreases from at least about 1600 ppm to less than about 1200 ppm. Referring to FIG. 10A which is a schematic representation of a cross-sectional view of a nuclear fuel rod for a BWR such as shown in FIGS. 6-8 but instead has a composite cladding 160 comprising an outer layer 161 and an inner layer 164 each of which is composed of a zirconium and or zirconium alloy metal. Outer layer 161 has a higher oxygen content than inner layer 164 (which is depicted as different stippling) so that hydrides will preferentially precipitate in the inner layer 164 and be inhibited from precipitating in outer layer 161. In a preferred embodiment, outer layer 161 has an oxygen content greater than about 1600 ppm, and inner layer 164 has an oxygen content less than about 1200 ppm. Referring to FIG. 10B which is a schematic representation of a cross-sectional view of another nuclear fuel rod for a BWR such as shown in FIGS. 6-8 but instead has a composite cladding 170 comprising an outer layer 171, an inner layer 174 and an innermost layer 177. Outer layer 171 and inner layer 174 are composed of a zirconium and or zirconium alloy metal. Outer layer 171 has a higher oxygen content than inner layer 174 (which is depicted as different stippling) so that hydrides will preferentially precipitate in the inner layer 174 and be inhibited from precipitating in outer layer 171. Innermost layer 177 can be zirconium or a zirconium alloy, or another metal. In a preferred embodiment, outer layer 171 has an oxygen content greater than about 1600 ppm, and inner layer 174 has an oxygen content less than about 1200 ppm. Referring to FIG. 10C which is a schematic representation of a cross-sectional view of another nuclear fuel rod for a BWR such as shown in FIGS. 6-8 but instead has a composite cladding 180 comprising an outer layer 181, an inner layer 184, and an innermost layer 187 each of which is composed of a zirconium and or zirconium alloy metal. Outer layer 181 has a higher oxygen content than inner layer 184 (which is depicted as different stippling) so that hydrides will preferentially precipitate in the inner layer 184 and be inhibited from precipitating in outer layer 181. In a preferred embodiment, outer layer 181 has an oxygen content greater than about 1600 ppm, and inner layer 184 has an oxygen content less than about 1200 ppm. In another preferred embodiment, innermost layer 187 has an oxygen concentration which is at least that of inner layer 184 but less than or equal to outer layer 181. Referring to FIG. 11 which is a schematic representation of a cross-sectional view of another nuclear fuel rod for a BWR such as shown in FIGS. 6-8 but instead has a composite cladding 190 which comprises at least two layers of zirconium and/or zirconium alloy metals, including an outermost layer 191 and an innermost layer 197. Outermost layer 191 has a decreasing oxygen concentration gradient depicted as a decreasing variation in stippling from outer wall 192 to an inner portion 191a where despite a temperature gradient across the cladding wall, zirconium hydrides will preferentially precipitate in inner portion 191a of outermost layer 191 and will be preferentially inhibited from precipitating near outer wall 192 of outermost layer 191. Although innermost layer 197 can be zirconium metal or alloy, or another metal or alloy, in a preferred embodiment, innermost layer 197 is a zirconium metal or alloy, and the oxygen content in outermost layer 191 of cladding 190 decreases from an amount greater than about 1600 ppm at the outer wall 192 to less than about 1200 ppm in the inner portion 191. While the present invention has been particularly shown and described with reference 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.
	claims	1. An electron beam exposure apparatus that has the ability to expose patterns corresponding to the selected aperture patterns, at one time, comprising an electron gun that generates an electron beam, a block mask equipped with plural aperture patterns, mask deflectors to deflect the said electron beam so that it passes selectively through one of the said plural aperture patterns, and also to deflect the said passed electron beam so that it returns to its original path, convergent devices that converge the said electron beam that has passed through the said block mask onto a specimen, and deflectors that deflect the said electron beam on the said specimen, wherein: the said plural aperture patterns are square or rectangular and arranged in a matrix form, and each aperture pattern has a square or rectangular maximum aperture area that limits the area in which each aperture is formed; the said block mask having at least one adjusting aperture pattern equipped with independent apertures of the same shape arranged along the opposite sides of the said maximum aperture area is provided; and the said mask deflector is adjusted so that the intensity of the beam, which is radiated onto the said specimen, at the portion of the independent apertures of the same shape arranged along the opposite sides of the said adjusting aperture pattern, is uniform and maximum; wherein, the electron beam entering onto the block mask has a size covering the maximum aperture area and passes simultaneously through each of the independent apertures. 2. An electron beam exposure apparatus as set forth in claim 1 , wherein, the intensity of the beam, which is radiated onto the said specimen, at the portion of the said apertures of the said adjusting aperture patterns, is measured by deflecting the electron beam shaped into the said adjusting aperture patterns using the said deflection means and by detecting the reflected electron when the fine lines on the specimen, along the sides of the said maximum aperture area, are scanned. claim 1 3. An adjusting method of an electron beam exposure apparatus, which can be used in an electron beam exposure apparatus that has the ability to expose the patterns corresponding to the selected aperture patterns at one time, comprising an electron gun that generates an electron beam, a block mask equipped with plural aperture patterns, mask deflectors to deflect the said electron beam so that it passes selectively through one of the said plural aperture patterns, and also to deflect the said passed electron beam so that it returns to its original path, convergent devices that converge the said electron beam that has passed through the said block mask onto a specimen, and deflectors that deflect the said electron beam on the said specimen, wherein: the said plural aperture patterns are square or rectangular and arranged in a matrix form, and each aperture pattern has a square or rectangular maximum aperture area that limits the area in which each aperture is formed; and the adjustment of the said mask deflectors is carried out so that the block mask having at least one adjusting aperture pattern equipped with independent apertures of the same shape arranged along the opposite sides of the said maximum aperture area is provided and that the intensity of the beam, which is radiated onto the said specimen, at the portion of the apertures of the same shape arranged along the opposite sides of the said adjusting aperture patterns, is adjusted so as to be uniform and maximum; wherein, the electron beam entering onto the block mask has a size covering the maximum aperture area and passes simultaneously through each of the independent apertures. 4. An adjusting method of an electron beam exposure apparatus as set forth in claim 3 , wherein the intensity of the beam, which is radiated onto the said specimen, at the portion of the apertures of the said adjusting aperture patterns is measured by deflecting the electron beam shaped into the said adjusting aperture patterns using the said deflection means and by detecting the reflected electrons when the fine lines, on the specimen along the sides of the said maximum aperture, area are scanned. claim 3 5. An adjusting method of an electron beam exposure apparatus as set forth in claim 3 , wherein: claim 3 the said block mask consists only of the said adjusting aperture patterns; and the adjustment of the said mask deflectors is carried out on all of the said adjusting aperture patterns. 6. An adjusting method of an electron beam exposure apparatus as set forth in claim 3 , wherein: claim 3 the said adjusting aperture patterns are arranged in the central and peripheral portions of the plural aperture patterns of the said block mask; and the adjustment of the said mask deflectors is carried out for the said plural adjusting aperture patterns.
	summary
053923243	description	DESCRIPTION OF A PREFERRED EMBODIMENT FIG. 1 shows a part of the secondary cooling system 1 of an integrated-type fast-neutron nuclear reactor. The nuclear reactor includes a vessel 2 fixed to the concrete structure 3 of the nuclear reactor and containing primary-coolant liquid sodium up to a level 4. The reactor core 5 constituted by Juxtaposed fuel assemblies is submerged inside the liquid sodium filling the vessel 2. Intermediate heat exchangers 6 pass through the slab for closing the upper part (not shown) of the vessel 2 and are submerged over a part of their height in the liquid sodium filling the vessel. In the nuclear-reactor vessel 2 or primary vessel are placed structures constituting the internal vessel making it possible to channel the circulation of the primary liquid sodium cooling the core 5. The intermediate heat exchangers 6 are placed, in relation to the internal vessel, so that the high-temperature primary sodium leaving the core penetrates into the intermediate heat exchanger via an upper port 6a and emerges from the intermediate heat exchanger 6 at a lower temperature, via a lower port 6b. The cooled sodium leaving the intermediate heat exchangers is circulated by pumps which enable it to be discharged to the lower part of the core 5. The intermediate heat exchangers include a heat-exchange bundle which is connected, for feeding it with secondary sodium, to a feed pipe 8 of the secondary cooling system 1. The secondary sodium circulates in the heat-exchange bundle of the intermediate heat exchanger 6, cools the primary sodium circulating inside the intermediate heat exchanger and is heated by thermal contact with the primary sodium, before emerging from the intermediate heat exchanger 6 via the pipe 9 of the secondary cooling system. The feed pipe 8 is connected to the discharge part of a secondary pump 10, the intake part of which is connected via a pipe 12 to the widened lower part 13 of the casing 15a of the steam generator 15. The outlet pipe 9 of the secondary cooling system is connected to the casing 15a of the steam generator at the widened upper part 14 of this casing. Beneath the widened lower part 13, the steam generator includes a water box 16 and, above the widened upper part 14, a steam manifold 17. The circulation of the secondary coolant, constituted by liquid sodium, inside the steam generator, is regulated so that the upper level of the sodium in the casing of the steam generator 15 lies in a substantially fixed position inside the widened upper part 14 during the operation of the secondary cooling system. The structure of the steam generator 15 will now be described in more detail with reference to FIG. 2. The steam generator 15, which is a generator of the type having straight tubes, includes a bundle 20 of long-length rectilinear tubes placed parallel to a direction which corresponds to the longitudinal direction of the bundle 20 and of the steam generator 15. The tubes of the bundle are welded, at one of their ends, to the tube plate of the water box 16 and, at their other end, to the tube plate of the steam manifold 17. In operation, the steam generator, as represented in FIGS. 1 and 2, is placed vertically in the structure of the nuclear reactor, i.e. in such a way that the longitudinal direction of the bundle 20 and of the steam generator 15 is vertical. The casing 15a of the steam generator, in its central part between the widened parts 13 and 14, has the shape of a cylindrical tubular casing containing the central part of the bundle 20. The secondary liquid sodium is heated in an intermediate heat exchanger by thermal contact with the primary liquid sodium, then transported via the pipe 9 and injected into the widened upper part 14 of the casing of the steam generator at a nozzle 9a. The secondary liquid sodium flows in the vertical direction from the top down inside the casing of the steam generator, and emerges from this casing via a nozzle 12a connected to the pipe 12 on which the circulation pump 10 is placed. Inside the casing of the steam generator 15, the liquid sodium circulates in contact with the outer surface of the tubes of the bundle 20 which are each connected, at one of their ends, to the water box 16 and, at their other end, to the steam manifold 17. When the nuclear reactor is operating, feed water is injected into the water box 16, which distributes the water into the tubes of the bundle 20. The water flows from the bottom up in the tubes of the bundle, is heated inside the tubes by thermal contact with the secondary liquid sodium, and is then vaporized and recovered in the form of steam in the upper manifold 17. According to the method of the invention, the nuclear reactor steam generators, such as the steam generator 15, may be used to remove the residual energy from the nuclear reactor, in the form of heat, after a shutdown by inserting the absorbing rods in the maximum insertion position into the assemblies for controlling the core. In order to cool the reactor by the method of the invention, either at shutdown or under accident operating conditions, each of the steam generators, such as the steam generator 15, is associated with a cooling device 22, such as represented in FIG. 2. The cooling device 22 includes a unit 25 of tubular shape which is placed around central part of the casing 15a of the steam generator 15, a chimney 24 and a pipe 23 linking the tubular unit of the device 22 to the chimney 24. The upper part of tubular unit 25 is fixed to the reactor structure supporting the steam generator 15 by means of a bellows Joint 26 fixed beneath an opening for the casing 15a of the steam generator 15 to pass through. At its other end (not shown in FIG. 2), the tubular unit 25 is connected in a sealed manner to the casing 15a of the steam generator 15. The lower part of tubular unit 25 includes air inlet openings 27 at which air passage ports 28, equipped with dampers for regulating the air flow rate, are fixed. A bellows joint 29 is interposed on the pipe 23 joining the tubular part 25 of the cooling device 22 to the chimney 24. A damper unit 30 makes it possible to regulate the air flow rate in the pipe 23, upstream of the chimney 24. The bellows, such as 26 and 29, make it possible to absorb differential deformations due to thermal expansion between the elements of the cooling device 22 and the nuclear reactor support structure. The chimney 24, the height of which may be considerable (for example twenty five meters in the case of a steam generator of a fast-neutron nuclear reactor of the currently constructed type) makes it possible to create an air inrush and a flow of air from the bottom up, inside the tubular unit 25 of the device 22, via the pipe 23. The air penetrates the lower part of the tubular unit 25 via the ports 28 and flows from the bottom up in the annular space between the uit 25 of the device 22 and the casing 15a of the steam generator. The flow rate of air entering the tubular unit 25 (in the direction of arrow 31) may be regulated by the dampers equipping the ports 28. Likewise, the air flow rate may be regulated at the outlet of the tubular unit 25 and at the inlet of the chimney 24 by the dampers 30. The flow of cooling air from the bottom up in the annular space between the tubular unit 25 and the external surface of the casing 15a of the steam generator 15 may be provided solely by draught and by natural convection or may be established, under forced conditions, by using a fan. Reference will now be made to FIGS. 3 and 4 in order to describe in more detail the structure of the tubular unit 25 surrounding the casing 15a of the steam generator 15. The tubular unit 25 includes a metal shell 32 placed coaxially with the casing 15a of the steam generator and connected at its upper part to the reactor structure via the bellows 26. The shell 32 is covered by a sleeve 33 made of thermally insulating material which thermally insulates the outer surface of the shell 32 from the outside environment. On the inner surface of the shell 32 are fixed, for example by welding, fins 34 constituted by plane portions of metal sheet, for the purpose of increasing the surface area for heat exchange by radiation. The fins 34 are fixed to the shell 32, in the longitudinal direction, along generatrices of this shell. Successive sets of fins, uniformly spaced along the internal periphery of the shell 32, as represented in FIG. 3, are placed along the length of the shell 32. Reference will now be made to all the figures in order to describe the implementation of the method for removing residual energy according to the invention, by removal of heat via the secondary cooling system, during a reactor shutdown. The reactor is shut down by inserting elements creating very strong negative reactivity in the reactor core 5. In order to remove the residual power from the nuclear reactor after shutting it down, the secondary liquid sodium continues to be circulated in the secondary cooling system 1 by the secondary pumps such as the pump 10. The secondary sodium thus circulates inside the intermediate heat exchanger 6 and cools the primary sodium which recovers the residual energy from the reactor, as manifested by a release of heat inside the core. The secondary sodium recovered at the outlet of the intermediate heat exchanger 6 via the pipe 9 is sent into the upper part of the steam generator 15. The secondary sodium then circulates inside the casing of the steam generator 15 and emerges at the lower part of the casing, via the pipe 12. During the cooling of the reactor at shut-down, the water feed of the water box 16 of the steam generator 15 is interrupted and the dampers of the ports 28 and of the pipe 23 are opened so that cooling air flows, from the bottom up, in the annular space 35 between the shell 32 and the casing 15a of the steam generator. The heat of the secondary liquid sodium circulating in the casing 15a of the steam generator is transmitted through the wall of the casing 15a by conduction and then from the outer surface of the casing 15a by radiation, so that the heat flux flows between the casing 15a of the steam generator and the shell 32, either directly (arrows 37) or via the fins 34 (arrows 38). The heat flux also flows between the fins 34 (arrows 39). The fins 34 channel the cooling air flowing in the annular space 35 and removing the heat transmitted by the casing 15a of the steam generator and coming from the secondary sodium. The cooling device 22, in particular because of its structure represented in FIGS. 3 and 4, makes it possible to remove efficiently the heat transported by the secondary sodium inside the casing of the steam generator. It is therefore possible to take up the residual heat from the nuclear reactor very efficiently, it being possible for the air flow rate in the device 22 and the thermal power of this cooling device 22 to be very high. Furthermore, the cooling device 22 may operate without energy from outside, the cooling air flowing by natural convection inside the annular space 35. If operation under forced-convection conditions is envisaged, an additional fan can be provided, but this is not necessary in most cases. The method and the device according to the invention therefore make it possible, very simply and very efficiently, to cool the nuclear reactor at shutdown and to remove its residual energy. The invention only implements devices and cooling systems existing in the nuclear reactor, or very simple additional devices such as the cooling device 22. Furthermore, the structure of the steam generator is not modified, the cooling device being fixed to the support structure of the steam generator and including no element fixed to the steam generator itself. It is possible to use a cooling device other than the one which has been described and to use a heat-exchange gas other than atmospheric air. Finally, the device and the method according to the invention may be used to remove residual energy from a nuclear reactor which includes steam generators of a type other than the straight-tube type.
052727378	abstract	There is disclosed a nuclear reactor installation. An accumulator-type emergency core cooling system, a gravity-driven core cooling system and an equalizing system for submerging a reactor core are provided within a primary containment vessel containing a reactor pressure vessel in which the reactor core is disposed. These cooling systems are automatically operated sequentially in accordance with the pressure in the reactor pressure vessel without the need for any particular powered source. The primary containment vessel is made of steel, and the interior of this containment vessel is divided into a space containing the reactor pressure vessel and a space containing an operation floor in such a manner that the two spaces are isolated from each other. An opening device is provided between the two spaces so as to communicate the two spaces with each other in the event of an accident, the upper portion of the primary containment vessel being air-cooled while the lower portion of this containment vessel being water-cooled.
	abstract	Disclosed herein is a method comprising coating a fissile, uranium-containing ceramic material with a water-resistant layer, the layer being non-reactive with the fissile, uranium-containing ceramic material. The coating is applied to a surface of the fissile, uranium-containing ceramic material. Also disclosed is a fuel for use in a nuclear reactor.
052727416	summary	BACKGROUND OF THE INVENTION The present invention relates to a nuclear fuel assembly and, more particularly, to a nuclear fuel assembly for a boiling water reactor having spacer structure improved on heat transfer from fuel rods to the coolant. Fuel assemblies for nuclear reactors have been improved on spacer structures thereof to increase heat transfer from the fuel rods to the coolant. In a case of a pressurized water reactor (PWR) as disclosed in U.S. Pat. Nos. 3,395,077 and No. 3,379,619, spacer structures, used for holding the fuel rods to keep them laterally spaced from each other, are directed to improve the heat transfer. The structure is such that, in the center of four sides surrounding fuel rods, a grating type spacer and an obstacle serving as a vane are provided. In this structure, the coolant flows over the peripheries of the fuel rods so as to cover them due to suitable configuration and mounting positions of the vanes. As a result, the coolant is agitated, the heat transfer increases, thereby raising allowable power level of a reactor core. The above-described art relates to a PWR. If the spacer structure is used in a boiling water reactor, it is impossible to achieve the above-mentioned effect. In such cases, voids occur in the core of the BWR, and the coolant flows in a two-phase flow. Namely, a liquid film flow takes place on the surface of the fuel rod and a mixture of steam and liquid drops flow in a space region enclosed by fuel rods. Under this two-phase flow condition, if the above-mentioned conventional spacer is used, the coolant flows along the periphery of each fuel rod. Such a flow strips off the liquid film adhered to the fuel rod, thereby decreasing the amount of liquid adhered to the fuel rod in the form of liquid film. Namely, such a coolant flow is likely to cause boiling transition that nucleus boiling changes to film boiling. Therefore, power, that is, allowable power level at the boiling transition decreases. An example of a BWR fuel assembly having a plurality of vanes is disclosed in U.S. Pat. No. 4,698,204. The U.S. patent relates to a BWR fuel assembly having an intermediate flow mixing non-support grid. The grid does not support fuel rods to keep them spaced from each other and having vanes at all the fuel rod for mixing relatively cool coolant and relatively hot coolant. SUMMARY OF THE INVENTION An object of the invention is to provide a nuclear fuel assembly which has sufficiently high allowable power level and stability. Another object of the invention is to provide a nuclear fuel assembly which has sufficiently high allowable power level without causing an increase in pressure loss. The present invention resides in a nuclear fuel assembly for a BWR comprising a plurality of fuel rods, a polygonal channel box surrounding the fuel rods, a plurality of spacers axially spaced from each other and each keeping the fuel rods laterally spaced from each other, and a plurality of vanes disposed only in a region at and around a corner within the channel box, for generating swirling flows in the region to thicken a liquid film on each fuel rod in the region. According to an aspect of the present invention, the spacers are of round-cell type, and constructed of a plurality of cylindrical cells joined each other, and at least one of the spacers which is disposed in an upper region in which boiling transition is likely to occur comprises vane-formed cylindrical cells having vanes formed on the outer surface thereof and smooth-surface cylindrical cells having no such vanes as mentioned above, the vane-formed cells being disposed, in use, for holding fuel rods in a farther region from a control rod, and the smooth-surface cells being for holding fuel rods in a closer region to the control rod, whereby the thickness of liquid films on the fuel rod in the farther region is thickened and heat transfer from the fuel rods to the coolant increases effectively. When the above-mentioned vanes are disposed in two phase flow region in the nuclear fuel assembly of the BWR, the vanes generate swirling flows of steam and liquid drops in spaces enclosed by adjacent opposite fuel rods. The liquid drops in the steam are moved to the fuel rods by centrifugal force due to the swirling movement of the liquid drops and adhered to liquid films on the fuel rods, thereby to increase the thickness of the liquid film flows on the fuel rods. Therefore, thermal allowance to boiling transition is improved and the allowable power level increases. The thermal allowance increases by increasing more the intensity of the swirling movement of the coolant. The increase in the intensity of the swirling movement can be achieved by making the vanes larger in scale, however, the vanes of large scale cause an increase in pressure loss. It is necessary to increase the thermal allowance without increasing the pressure loss. The boiling transition does not occur at all the fuel rods, but it takes place locally. In general, the position where the boiling transition is likely to occur is one that power is high and thermal conditions are severe. That position is a corner at which fuel rods are positioned and which is farthest from a control rod and in the vicinity of the corner. A typical example of the corner is defined by two sides of the channel box and farthest from the corner defined by other two sides of the channel box which face two sides of a cruciform control rod. Therefore, when the spacer is constructed so that swirling flows take place only in spaces enclosed by adjacent opposite fuel rods and in spaces between channel box and the fuel rods facing the channel box in the corner region as mentioned above, the thermal allowance is increased without causing pressure loss since the number of vanes used there is small, even if vanes of large scale are employed. Use of the vanes of large scale increases an amount of liquid film flow flowing along the surfaces of the fuel rods. When a large number of vanes are used, projection area of the spacer increases, so that an area of flow passage defined within the spacer is made smaller, as a result, the pressure loss increases. In the present invention, the number of vanes can be reduced to 15 or less, for example, so that the pressure loss little increases. As mentioned above, the space where the boiling transition is apt to occur is at and around the corner of the insides of the spacer further from the control rod in a lateral plane. With respect to a vertical position, the boiling transition takes place at the upstream sides of first and second stage spacers from the top side. Usually, a fuel assembly comprises around 7 stage spacers. The boiling transition is apt to occur around the highest stage spacer. The liquid film thickness decreases from the lowest stage toward the highest stage because the liquid film is evaporated. However, at the spacers, the liquid film increase because the coolant is agitated by the spacers. An aspect of the invention is in that the vanes are provided on the second and third spacers, so that the thickness of liquid film on the fuel rods on which the liquid film is thin increases at the upstream sides of the spacers.
	description	This application is a divisional application of and claims priority under 35 U.S.C. §§120, 121 to co-pending application Ser. No. 11/002,680 “Method of producing isotopes in power nuclear reactors,” filed Dec. 13, 2004, the entirety of which is incorporated by reference. A wide variety of radioactive isotopes are used for medical, industrial, research, and commercial applications. In general, radioisotopes may be produced by irradiating target isotope materials with nuclear particles. The target atoms either transmute directly into the desired isotope, or a radioisotope is produced through a chain of absorption and decay that subsequently generates the desired radioactive product. The latter is the case, for example, in the production of the important medical imaging isotope Technetium-99m, which represents about 90% of the isotopes used in today's nuclear medicine. Tc-99m produces high-energy gamma rays, which makes detection easy, and its short half-life of 6 hours minimizes the radiation dose delivered to the patient. Nuclear medical imaging is unique in its ability to analyze organ structure and functioning. It allows the gathering of diagnostic information that would otherwise require surgery, or not be available at all. It also promotes early detection and treatment of cancers and other problems. Two different technologies are used to provide the source of radiation for radioisotope production: nuclear reactors, which produce a flux of neutrons, and particle accelerators or cyclotrons, which produce a flux of charged particles, usually protons, but sometimes electrons or other particles. As an example, Tc-99m is the decay daughter product of Mo-99, an isotope with a half-life of 66 hours that is produced in nuclear reactors as a fission product of the neutron bombardment of a uranium target material. This is the source of all of the Tc-99m used in the world today. After irradiation, Mo-99 is recovered from the target, and packed into a production device from which technetium may be eluted in substantially pure form as needed for preparing radiopharmaceuticals for nuclear medical procedures Other exemplary radioisotopes used for medical, industrial, research and commercial applications include thallium-201, which is used for medical cardiac imaging; calcium-44, which is used in bone growth studies; iridium-192, which is used for nondestructive testing of construction and other materials; cobalt-60, which is used to destroy cancer cells, to disinfect surgical equipment and medications, and the sterilization of food supplies; thulium-170, which is used for portable blood irradiations for leukemia, lymphoma treatment, and power source; gadolinium-153, which is used for osteoporosis detection and SPECT imaging; nickel-63, which can be used for the creation of long-life batteries; and americium-241, which is used in smoke detectors. In addition, rapid advances in nuclear medicine and other fields are focusing attention on a number of isotopes that have not previously been considered commercially important. As an example of a conventional reactor used in radioisotope production, medical and industrial radioisotopes have been produced since 1957 by Canada's National Research Universal (NRU) reactor at the Atomic Energy of Canada's (AECL's) Chalk River Laboratories in Ontario, Canada. The NRU is a CANDU reactor designed for neutron beam research, materials research and medical/industrial isotope production. In general, CANDU plants are designed to use heavy water (deuterium oxide, or D2O) as the moderator and coolant. The use of heavy water, combined with other features of its design, allows the CANDU reactor to use natural uranium (NU) fuel, which is relatively inexpensive and abundant worldwide. The NRU produces a high percentage of the world's medical and industrial radioisotopes, including molybdenum-99, a critical isotope used for medical diagnoses. In general, specimen rods containing an isotope target are inserted through penetrations in the NRU in a continuous process and subject to irradiation therein, so as to produce isotopes at a desired specific activity for use in nuclear medicine and/or industrial applications. Two specialty reactors, the MAPLE 1 and MAPLE 2, are under construction at Chalk River Laboratories. These reactors are intended to replace the NRU. The MAPLE 1 and MAPLE 2 are dedicated exclusively for the production of medical radioisotopes. These research reactors are not intended nor designed for commercial power generation, since they are being designed for power levels of about 10 MWt. The MAPLE is a low-pressure, low-temperature, open-tank-in-pool type research reactor that uses low-enriched uranium (LEU) fuel. The core is compact, and is cooled and moderated by light water. Surrounding the light water core is a heavy water reflector tank, which maximizes the available neutron fluxes needed for radioisotope generation. Exemplary embodiments of the present invention, in general, may be directed to a method of producing isotopes in a light water power reactor, in which at least one or more targets within the reactor is irradiated under a neutron flux to produce at least one or more isotopes. The targets may be assembled into one or more fuel bundles that are to be loaded in a core of the reactor at a given outage. Power operations in the reactor irradiate the fuel bundles so as to generate desired isotopes, such as one or more radioisotopes at a desired specific activity or stable isotopes at a desired concentration. As to be described in further detail below, the exemplary embodiments may be directed to using power light water reactors designed for commercial power generation to produce medical and industrial isotopes, in addition to generating thermal and electrical power for residential, commercial and/or industrial use. Unlike research reactors dedicated only to the production of radioisotopes, these light water power reactors may be existing boiling water reactors (BWRs) and/or pressurized water reactors (PWRs) that are configured to generate at least 100 megawatts thermal (MWt) and in example on the order of several hundred or thousand MWt. These power LWRs generally have a bundle average enrichment of less than 10%. FIG. 1 is a flow chart illustrating a method of producing isotopes in a light water power reactor, in accordance with an exemplary embodiment of the invention. Referring to FIG. 1, isotope target material may be loaded into a light water power reactor during a planned outage period between power operations or even prior to initial reactor operation (S100). In an example, the targets may be loaded in one or more fuel bundles to be assembled in the reactor during the outage period. In a further example as will be seen in further detail below, the exemplary fuel bundle may be configured to include a plurality of one or more (or all) full length rods, one or more (or all) part length rods, one or more rods composed of a plurality of segments to realize a segmented rod, or combinations of one or more full length, part length and segmented rods. In another aspect, the isotope target may be housed within a containment structure that is inserted or placed within one or more given full length rods, part length rods, segmented rods or water rods of the fuel bundle so as to form or “target rods” for irradiation in the bundle of the reactor. Once the fuel bundle(s) are assembled with one or more target rods, the fuel bundle(s), which may be occasionally referred to as “bulk isotope generation bundles (BIGs)” may be assembled in the core during the outage period. Thereafter, power operations for a next energy cycle in the reactor may be commenced to irradiate the target material therein (S200). The energy cycle represents a period of time (an exposure period as measured in megawatt-days per short time (MWD/st)) in which the reactor is operated for generating power between planned outages. Exemplary energy cycles for BWRs and/or PWRs may be 12 months, 18 months, 2 years, etc., for example. During power operations, those fuel bundles containing targets are irradiated under a neutron flux in order to produce desired isotopes. The produced isotope may be understood as a radioisotope produced from the target isotope material at a desired specific activity, or a stable isotope that is produced from the target isotope material at a desired concentration, as the target absorbs one or more neutrons during the energy cycle. The produced isotopes may be removed at a pre-planned outage at end of cycle (S300) so that isotope product removal coincides with the pre-planned outage, so as to minimize the impact on the expect energy production of the light water power reactor, for example. Accordingly, at end of cycle, the reactor may be shutdown and those bundles with produced isotopes (BIG bundles) may be removed or discharged from the reactor during the planned outage following shutdown. FIG. 2 illustrates an exemplary fuel bundle of a light water power reactor such as a boiling water reactor (BWR). Fuel bundle 10 may represent an exemplary BIG bundle for producing isotopes in a light water power reactor, it being understood that the exemplary embodiments are not limited to the exemplary fuel bundle configuration in FIG. 2. Fuel bundle 10 may include an outer channel 12 surrounding an upper tie plate 14 and a lower tie plate 16. A plurality of rods 18a-c may be disposed in a matrix within the fuel bundle 10 and pass through a plurality of spacers (also known as spacer grids) 20 vertically spaced one from the other maintaining the rods 18a-c in the given matrix thereof. The rods 18a-c and at least a pair of water rods 22 and 24 may be maintained in spaced relation to each other in the fuel bundle 10 by a plurality of spacers 20 provided at different axial locations in the fuel bundle 10 so as to define passages for reactor coolant flow between the rods 18a-c in the fuel bundle 10. In FIG. 2, the matrix array of rods may be a 10×10 array, although the illustrative fuel bundle 10 may have a different matrix array of rods 18a-c, such as a 9×9 array. The bundle 10 may include all full length fuel rods 18a, all part-length rods 18b, all segmented rods 18c and/or any combination of full, part-length and/or segmented rods 18a-c; bundle 10 of FIG. 2 is shown including all configurations of rods 18a-c merely as an example. The water rods 22 and 24 (two are shown, there may be greater or fewer water rods in bundle 10) may be disposed among the rods 18a-c in bundle 10, between the lower tie plate 16 and the upper tie plate 14. The water rods 22, 24 serve to transfer moderator fluid from the lower regions of the nuclear fuel bundle 10 to the upper regions. FIG. 3 illustrates a tubing portion of a rod to illustrate a containment structure with target in accordance with an exemplary embodiment of the present invention. FIG. 3 shows a tubing portion 180 of an exemplary rod 18a-c of the fuel bundle 10, it being understood that this tubing portion 180 could be part of a full length rod 18a, part-length rod 18b and/or part of a segment of a segmented rod 18c. In an exemplary embodiment of the present invention, various ones of the rods 18a-c may include a container assembly therein. FIG. 3 illustrates a tubing portion 180, which may be a section of one of the rods 18a-c, for example, or alternatively, part of an upper or lower end plug. The tubing potion 180 is shown in phantom (transparent) so as to illustrate a container assembly 300 and various features or components of the container assembly 300 provided therein in more detail. Container assembly 300 may include a containment structure 310 housing one or more irradiation targets 320 that produce one or more desired isotopes when a fuel bundle 10 containing one or more rods 18a-c with container assembly 300 is irradiated in the core of the light water power reactor. As desired above, a rod 18a-c with the container assembly 300 may be referred to as a target rod. One or more tubing portions 180 of a given target rod 18a-c may each include the same, different or multiple irradiation targets 320, for example. Referring to FIG. 3, in one example the containment structure 310 may be substantially hollow, of a generally cylindrical shape and may be closed at one end 311, open at the other end 312 and may include a seal 313 to close the structure 310 by a suitable end cap 330, although end caps 330 may be provided at both ends. Although containment structure 310 is shown as having a generally cylindrical shape, containment structure 310 may be oriented in any geometrical shape so long as the largest diameter of the shape is less than the inner diameter of a tubing portion 180 of a given rod 18a-c. Containment structure 310 may be made of a suitable material such as a zirconium alloy, for example, although not limited to this material, it could also be a stainless steel or allow thereof. Containment structure 310 may house one or more irradiation targets (i.e., isotope targets) 320 therein. Target 320 shown in FIG. 3 is illustrated in a generally solid, cylindrical form or shape. However, the irradiation target 320 may be embodied as a solid, liquid and/or gas. In another example, the target 320 may be in a ‘BB’ form or may assume a pellet-like shape that is comparable to the pellet size of existing UO2 pellets, for example. Target 320 may take any geometry so long as the diameter of the geometry is small enough to fit inside the containment structure 310 (less than an inner diameter of the containment structure 310) within a given rod 18a-c. The containment structure 310 therefore provides a double containment for the isotope target 320 within the rod 18a-c when inserted therein at time of rod 18a-c manufacture. Optionally, an interior of the containment structure 310 may include a spring 340 to provide a counter force against target 320 when sealed by end plug 330. The end plug 330 may be attached to the containment structure 310 by suitable attachment means, i.e., weld, threaded engagement, friction connection, etc. The containment structure 310 may include a pilot hole 303 at end 311 for removing the target(s) 320 after irradiation (which include isotope material therein). End 311 may include exterior threads 301 and an O-ring 302 that is used for sealing container assembly 300 when inserted into a piece of equipment. Pilot hole 303 may include interior threads to aid in the removal of container assembly 300 from a rod 18a-c. Further details of suitable container assemblies are described in the issued patent entitled “Rod Assembly for Nuclear Reactors”, U.S. Pat. No. 7,526,058, the entire contents of which are incorporated by reference herein. One or more tubing portions 180 of a given rod 18a-c may thus include container assemblies 300 with the same target 320, different targets or multiple targets that produce different isotopes at a desired specific activity (radioisotopes) and/or different stable isotopes at desired concentration, for example. In other words, a given target rod 18a-c can include a plurality of the same or different targets 320 spaced axially at different locations along the target rod 18a-c. In one exemplary aspect of the invention, at least one target rod includes one or more container assemblies 300 but includes no nuclear fuel (i.e., “nuclear fuel” may be understood as desired enrichments of uranium of UO2 pellets and/or gadolinium). In an aspect, the target 320 may be coated with a magnetic material to improve solubility and prevent release of target material therein into the reactor coolant. In another aspect, one or more of the target rods 18a-c (full-length, part-length and/or segmented rods) may also include desired enrichments of uranium and/or gadolinium. A given fuel bundle 10 may include various combinations of rods 18a-c that include nuclear fuel, rods 18a-c that include no nuclear fuel but include targets 320, and rods 18a-c which have enriched uranium and/or gadolinium zoned at certain axial locations with targets 320 placed at other axial locations within the target rod 18a-c along the length thereof. The axial and/or radial locations and concentrations of fuel and poisons may be based on the desired characteristics of the bundle 10 for a planned energy cycle in the light water power reactor, for example. In an example, a rod segment of the segmented rod 18a-c that includes a container assembly 300 could also include nuclear fuel, although adjacent rod segments could include nuclear fuel therein. The target 320 may be a target selected from a group of isotopes having an atomic number greater than 3. In an example, the target 320 may be selected from isotopes comprising one or more of cadmium, cobalt, iridium, molybdenum, nickel, thallium, and/or thulium isotope. Desirably, a given tubing portion 180 and/or container assembly 300 may include indicia or indicators thereon to indicate what target 320 is loaded therein, for example, and/or what isotope is to be produced from that target. As discussed above, the target 320 may be embodied in solid, liquid or gas form and placed inside a tubing portion 180 of a target rod 18a-c. It may be desirable to place the target 320 in the reactor at a location that maximizes the specific activity of the produced radioisotope/concentration of the produced stable isotope at end of cycle. The properties of the target isotope and produced radioisotopes or stable isotopes may factor in this determination. Accordingly, the target 320 may be placed at a location in the reactor that may be determined as a function of the absorption cross-section of the target isotope itself, and also as a function of the half-life of the produced isotope and the absorption cross-section of the produced isotope. These factors influence the axial and radial zoning of target material within the fuel bundle and more specifically within individual target rods 18a-c of a BIG bundle such as the fuel bundle 10, for example. The absorption cross section of the containment structure 310 housing the target 320 may also be a factor. For some isotopes (such as IR-192), higher neutron fluxes are not necessarily desirable due to the fact that the target may prematurely burn out. By using a containment structure 310 with a material such as stainless steel instead of zircaloy, the desired reduced neutron flux on the target may be achieved, resulting in a high specific activity isotope at the end of cycle without premature burnout of the produced isotope. The axial and radial locations within the core may be substantially strong independent variables that affect the amount and spectral characteristics of neutron flux that a target will observe. For example, if the absorption/decay characteristics indicate that a larger flux is desired to maximize specific activity, placement of the bundle away from the core boundaries (towards center of core) in both radial and axial directions may be desired. Similarly, if the absorption/decay characteristics indicate that a lower flux is desired to prevent burnout of the target isotope, then the target bundle should be placed towards the perimeter of the core and/or at the extreme top or bottom of the bundle. Also, selection of the location of the target rod within a given target BIG bundle can greatly influence the final specific activity. As shown in the bundle 10 example of FIG. 2, rods located towards the perimeter and edge of the bundle in a BWR result in a greater thermal flux than the rods towards the inner regions of the bundle. Therefore based on the absorption/decay characteristics, appropriate rod locations may be selected. Although the container assembly 300 with target 320 has been described as being located in a tubing portion 180 of a given rod 18a-c where there is no nuclear fuel, for some isotope targets it may be desirable to surround the container assembly 300 with additional enrichment. For example, a tubing portion 180 containing the containment structure 310 may be packed with additional enrichment in the area between the containment structure 310 outer surface and the inside of the tubing portion 180 during rod manufacture. By packing additional uranium in proximity to the target 320 within the containment structure 310, additional neutron flux may be obtained for the target 320. Consequently, the specific activity of the produced radioisotope may be higher and more valuable to the customer. In addition to adding enrichment, another factor which may affect the neutron fluxes surrounding the containment structure 310 may be the moderator. Packing a tubing portion 180 with additional or alternative moderator (such as graphite) in the space between the containment structure 310 and the inside of the tubing portion of a given rod 18a-c during rod manufacture may enable additional neutron flux to be obtained for the target 320 during power operations in the reactor. Done prior to reactor operation, providing additional moderator in closer proximity to a desired target location in the reactor may thus influence the energy spectrum of the neutron flux through the target during power operations. Heretofore, the exemplary embodiments have been described loading an isotope target 320 within a containment structure 310 to be provided in one or more of the rods 18a-c of the exemplary bundle 10. However, depending upon the properties of the target isotopes and that of the produced radioisotope or stable isotope (i.e., absorption cross section, half life, etc.), it may be desirable to place the target in an axial location within the fuel bundle 10 that typically receives less neutron flux during an energy cycle. For example, each of rods 18a-c (and water rods 22, 24) include an upper end plug and a lower end plug that are connected to upper and lower end tie plate 14 and 16 in the bundle 10. One or more container assemblies 300 may be placed within interior portions of one of the upper or lower end plugs during the rod manufacturing process, for insertion into the fuel bundle 10 at a planned outage. Alternatively, a containment structure housing one or more targets 320 therein may be placed within a control blade during a control blade manufacturing process. For example, a control blade which is used for reactor control between fuel bundles may have its neutron absorbers (i.e., B4C, Hf, etc.) replaced with a container assembly 300. Such control blades would not be used for reactor control and could be used as radioisotope target vehicles into and out of the core. This may allow for insertion and removal of targets 320 at times other than reloads at a planned outage, for example. This may also provide flexibility in axial elevation selection for neutron flux tuning within the core. A core of a light water power reactor includes several penetrations for insertion of instrumentation tubes during the energy cycle to take measurements, readings, etc. Accordingly, a containment structure 310 with radioisotope target 320 may be inserted into BWR instrumentation tubes through the use of a traversing and core probe machine, for example. This may also allow for insertion and removal of targets, to perhaps produce isotope material for those isotope having shorter half-lives or which do not need as much neutron flux to be removed at a time other than a planned reload at a planned outage. In another alternative, isotope targets may be placed outside the core but within the pressure vessel. A typical light water power reactor such as a BWR includes a core shroud which connects the core to the reactor pressure vessel. A containment structure 310 with isotope target 320 therein may be attached to the core shroud during a planned outage and subsequently irradiated during a next energy cycle for irradiating the target 320 to produce a desired isotope therein. Placing the target 320 outside the core may facilitate licensing since it does not include core internals such as fuel, control blades, etc. Targets 320 which produce isotopes that require low flux may be desirable candidates for placing on the core shroud and may withstand an entire cycle of the radiation, for example. FIG. 4 is a flow chart illustrating a method of producing isotopes in target rods of a fuel bundle within a light water power reactor, in accordance with an exemplary embodiment of the invention. FIG. 4 illustrates an exemplary process 400 for producing isotopes in a LWR. FIG. 4 illustrates exemplary process steps for assembling a bulk isotope generation bundle for insertion into the core of the light water power reactor during a planned outage, and also illustrates process steps for the removal of produced isotopes from the core and subsequent processing. The target rods 18a-c may be manufactured (S410) in advance. For example, desired isotope target materials, and target rod components may be procured in advance to manufacture target rods, which may be embodied as one or more of the rods 18a-c in FIG. 2. A given target rod may include a plurality of containment assemblies 300 inserted therein that include all of the same target, or include different targets located at desired axial locations in the rod to produce a radioisotope at a desired specific activity and/or desired stable isotope at a desired concentration, based on the properties of the target isotope material and/or known properties of the produced isotope. Additionally, carrier bundle uranium and components for the bulk isotope generation bundle may be procured (S420) and the bulk isotope generation (BIG) may be assembled from the target rods, procured uranium and associated components (S430). Standard reload bundle procurement and manufacture of the reload bundles (S440) may also be performed, as is known, to be loaded at next planned outage. The reload bundles and BIG bundles may be shipped to the reactor site for loading into the core of a desired light water power reactor such as a BWR. Reload licensing for the reload bundles would be done prior to shipping or prior to loading the bundles in the core. The same is expected for the BIG bundles. The BIG and reload bundles are loaded into the reactor core (S450). Up on completion of the outage, the reactor may commence power operations for a given energy cycle to irradiate the BIG bundles (S460). At end of cycle, the reactor is shut down and the BIG bundles may be discharged (S470) during the subsequent outage period to a staging area in the spent fuel pool. The staging area may be an enclosed work area or “safe chamber” within the spent fuel pool. The same chamber may be at a sufficient size to hold the BIG bundles and necessary tools, casks/shielded shipping containers and surveillance equipment. The safe chamber would allow for accidental dropping of target rods and prevent a potential radiation exposure accident and may provide a convenient way to secure and obtain quick access to surveillance equipment and disassembly tools. In the staging area, the target rods may be harvested (S480) from the BIG bundle using suitable tooling and loaded into shielded shipping containers or casks for transport to an off-site facility. Those harvested BIG bundles are then returned to the spent fuel pool. Alternatively, if the plant site includes a hot cell facility, the hot cell facility may be joined to the reactor by a suitable corridor through which the materials irradiated in the reactor may be brought to the hot cells (shielded work areas) where the isotopes may be extracted and packaged for shipment. Thus, the target rods are disassembled (S490) in the hot cell to retrieve the desired isotope materials for final processing and transport (S495) to the customer. Heretofore, the exemplary embodiments have been described in which the produced isotope is removed at a point in time to coincide with an end of cycle of a given energy cycle. However, there are occasions where an unplanned outage occurs at a point during the energy cycle, sometimes referred to as a mid-cycle outage. Accordingly, target isotopes may be loaded in the reactor that produce isotopes at a desired specific activity or stable isotope at a desired concentration at such a mid-cycle outage point. These radioisotopes or stable isotopes may then be removed at the mid-cycle outage point to provide a revenue stream that may potentially offset the costs associated with the unplanned outage period. Accordingly, the exemplary methodology may illustrate the advantages of assembling bulk isotope generation bundles for a core of an existing light water power reactor that is designed for commercial power operations. A BIG bundle may include one or more target rods, and each target rod may include one or more of the same or different target isotopes that, upon irradiation under a neutron flux during an energy cycle, produce radioisotopes at a desired specific activity or stable isotopes at a desired concentration for medical and/or industrial isotope applications. Since the core of the exemplary light water power reactor may include in upwards of several hundred fuel bundles, the potential to generate (bulk produce) medical and/or industrial isotopes to satisfy the world demand may be realized. As an example, a core loaded with no more than about 8 to 16 BIG bundles (out of a total of several hundred fuel bundles) could potentially produce sufficient isotope product to satisfy the current world demand for those applications that require medical and/or industrial isotopes. The exemplary embodiments of the present invention being thus described, it will be obvious that the same may be varied in many ways. Such variations are not to be regarded as departure from the spirit and scope of the exemplary embodiments of the present invention, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the following claims.
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044938110	summary	BACKGROUND OF THE INVENTION In general, a wide measuring range is required for a measuring system to measure the neutron flux level in a pressure vessel in a reactor. To measure a neutron flux level in a boiling water reactor (BWR), for example, the required measuring range is 11 decades. This implies that it is very difficult to obtain a satisfactory result by the use of a single measuring device. Conventionally, three measuring devices are combined to form a measuring system for neutron flux level measurement. Of these measuring devices, the first one is for low neutron flux level measurement. In the start-up range of the reactor operation, the neutron flux level is low and the reactor power is proportional to the counting rate of a neutron counter. The counting rate of the output pulses from a neutron counter capable of producing the pulse output up to 6 decades is sufficient for the neutron flux level measurement in this range. When the reactor operation enters the intermediate range between the start-up range and the power range, the neutron flux level increases to a high level. In this condition, the counting rate measuring method is insufficient for the reactor power measurement in this intermediate range. However, Compbell's theory is applicable for the reactor output measurement in this intermediate range, since the reactor output is proportional to a mean squared value of an AC component contained in the output signal of the neutron detector. Accordingly, in this range the neutron flux level can be measured using a measuring method based on Campbell's theory. In the power range, the reactor output is proportional to a DC component in the detector output signal. Accordingly, the reactor power in this range can be measured by measuring this DC component. For measuring the reactor power of the BWR, for example, 4 neutron detectors are installed for the start-up range; 8 detectors for the intermediate range; and 100 to 200 detectors for the power range. The types of neutron detectors are different for each reactor operation range. The measuring devices for the respective operation ranges will now be described in detail. FIG. 1 shows a neutron flux level measuring device for the start-up range. In FIG. 1, the output pulses of a neutron counter 1 installed in the pressure vessel are transferred to an input terminal of an input circuit 4 for a preamplifier 3 for pulse amplification, through a coaxial cable 2 which is 10 m to 20 m long. The input circuit 4 is a low input impedance circuit containing a capacitor C1 and a resistor R1. The input circuit 4 is impedance-matched to the coaxial cable 2 so as to prevent a signal reflection on the signal path therebetween. In FIG. 1, CO represents a capacitance of the coaxial cable 2, HV a high tension voltage terminal, and RH a resistor inserted between the high tension voltage terminal HV and the input terminal of the input circuit 4. FIG. 2 shows a prior art flux level measuring device for intermediate range measurement. In the figure, a neutron detector 5 for the intermediate range is coupled to an input terminal of an input circuit 7 for a preamplifier 6 for voltage amplification. The preamplifier 6 is of the low noise type, which amplifies an AC component output (Campbell signal) from the neutron detector 5. The input impedance of the input circuit 7 is selected to be from 5 kiloohms to 10 kiloohms. As already mentioned, it is technically very difficult to measure both the pulse signal and the Campbell signal by the use of the same measuring device. Specifically, it is very difficult to design a measuring device in which a pulse channel preamplifier and a Campbell channel preamplifier are connected to the same type of neutron detector for separating the signals from the neutron detector into a pulse signal and a Campbell signal and amplifying the corresponding signal so as to satisfy a required condition. It is also difficult to measure the reactor power so that the measuring range of the measuring device of FIG. 1 partially overlaps that of the measuring device of FIG. 2. A neutron flux measuring device shown in FIG. 3 is so designed as to amplify a pulse signal and a Campbell signal from a wide range neutron detector 8 by using a single preamplifier 9. The preamplifier 9 which is provided with a low input impedance circuit 10, including a capacitor C3 and a resistor R3, has low noise performance and a wide frequency band. In the measuring device shown in FIG. 3, a noise voltage e.sub.n (r.m.s.) which is converted into an input signal to the preamplifier 9 is expressed by equation (1) ##EQU1## where Rin is resistance of the resistor R3, k a Boltzman constant, 1.3804.times.10.sup.-25 Joul/.degree.K, T an absolute temperature (.degree.K), and B a frequency band width (Hz) of the preamplifier 9. For the noise voltage e.sub.n (r.m.s) as given above, a noise current i.sub.n (r.m.s), which is converted into an input current to the preamplifier 9 or an equivalent input noise current, is given by ##EQU2## Incidentally, the noise current is generated as a thermal noise from the resistor R3. The Campbell signal derived from the detector 8 is a current signal. Accordingly, if the Campbell signal from the wide range neutron detector 8 is smaller than the equivalent input noise current i.sub.n (r.m.s) as given by equation (2), it is impossible to detect the Campbell signal. Particularly, since the resistance Rin of the input resistor R3 of the preamplifier 9 is small (50 ohms or so), the equivalent input noise current i.sub.n (r.m.s) is extremely large. In the measurement of the Campbell signal using the measuring device shown in FIG. 3, the signal to noise ratio S/N is approximately 200 times that when the preamplifier 6 of FIG. 2, which has the input impedance circuit 10 including the resistor of 10 kiloohms, is used for measuring the same Campbell signal. For this reason, it is very difficult to amplify both the pulse signal and the Campbell signal by the single preamplifier 9 with a low input impedance. A measuring device shown in FIG. 4 is designed to measure a pulse signal and a Campbell signal by using one neutron detector, and is disclosed in Kokai (Japanese Unexamined patent publication) No. 56-117193. In the measuring device of FIG. 4, the neutron flux level signal detected by a fission counter 11 is applied through the coaxial cable 2 to a preamplifier 13 with an input impedance circuit 12 which is made up of a capacitor C4 and a resistor R4, and to a preamplifier 15 with an input impedance circuit 14 which is made up of a capacitor C5 and a resistor R5. The preamplifier 13 amplifies a high frequency band signal which is derived from the fission counter 11 and delivered through the input impedance circuit 12. The preamplifier 15 amplifies the intermediate frequency band signal which is derived from the fission counter 11 and delivered through the input impedance circuit 14. Thus, the preamplifier 13 is a pulse amplifier for amplifying a pulse signal, while the preamplifier 15 is a Campbell signal amplifier for amplifying a Campbell signal. In FIG. 4, the neutron flux level signal from the fission counter or the neutron detector 11 is routed to the input impedance circuit 12 or the input impedance circuit 14 depending on the frequency band of the signal. The preamplifier 13 amplifies only the high frequency component applied through the input impedance circuit 12. The preamplifier 15 amplifies only the intermediate frequency component, i.e., the Campbell signal, which is applied thereto through the input impedance circuit 14. Accordingly, the amplifiers 13 and 15 can amplify the neutron flux level signal with an improved S/N. The amplifiers 13 and 15 have frequency characteristics as shown in FIG. 5. In the graph, the x-distance represents frequency (Hz) and the y-distance an amplification factor (dB). In FIG. 5, f.sub.CO, f.sub.CL and f.sub.CH indicate respectively the center frequency, the lower limit frequency and the upper limit frequency of the Campbell signals. f.sub.PO, f.sub.PL and f.sub.PH respectively indicate the center frequency, the lower limit frequency and the upper limit frequency of the pulse signals. If the capacitance C4' of the capacitor C4 and the resistance R4' of the resistor R4 in the input impedance circuit 12 are selected so as to satisfy the following relations ##EQU3## the pulse signal component can be smoothly supplied to the preamplifier 13. If the capacitance C5' of the capacitor C5 and the resistance R5' of the resistor R5 of the input impedance circuit 14 are selected so as to satisfy the following relations ##EQU4## the Campbell signal component can be smoothly sent to the preamplifier 15. The noise applied from the preamplifier 13 to the preamplifier 15 can be removed by properly selecting the frequency characteristic of the preamplifier 15. A serious disadvantage of the measuring device shown in FIG. 4 is that it has a long response time. The response time means the time taken for the signal generated in the neutron detector 11 to be detected by a measuring system connected to the outputs of the preamplifiers 13 and 15. The response time of the measuring device of FIG. 4 will be described in detail referring to FIG. 6. The Campbell signal from the neutron detector 11 includes a fluctuation component .sigma..sub.s with respect to the level S of the Campbell signal. The fluctuation component is represented by a mean squared value of the amplitude of the Campbell signal. The ratio of the fluctuation component .sigma..sub.s to the Campbell signal level S, viz. .sigma..sub.s /S, is called a fluctuation factor, here denoted as I', and is given as ##EQU5## where .tau. is a response time, and Nn is a pulse rate (c.p.s) of the pulse signals derived from the neutron counter 11. For f.sub.CL and f.sub.CH, see FIG. 5. When f.sub.CL =1 kHz and f.sub.CH =10 kHz, and .tau. is 1 msec, 10 msec, 0.1 sec or 1 sec, the fluctuation factor I' of the Campbell signal versus the pulse rate Nn (c.p.s) is as shown in FIG. 6. As seen from FIG. 6, for measuring the neutron flux level in the intermediate range while keeping the fluctuation factor I' below approximately 1%, the response time must be 0.1 sec or more. However, in the case where the response time is 0.1 sec or more, the measuring devices can not effectively measure the neutron flux level within a very short response time when the reactor power abruptly increases and the pulse rate Nn is very large. The disadvantages of the prior measuring devices by which the pulse signal and the Campbell signal are measured by one neutron detector, will again be described as follows. If the measuring device shown in FIG. 3 is used for such a measurement, the S/N ratio is deteriorated in the measuring range of the Campbell signal. This leads to a narrow overlapping range of the Campbell signal measuring range and the pulse signal measuring range. As a result, the reliability of the measurement is deteriorated. In the measuring device shown in FIG. 4, the response time .tau. in the measuring range of the Campbell signals is made long. Therefore, the prior measuring device shown in FIG. 4 can not measure the neutron flux level when the neutron detector 11 produces pulse signals at high pulse rate and the measurement must be done within a short response time. SUMMARY OF THE INVENTION Accordingly, an object of the invention is to provide a preamplifier for a wide range neutron monitoring system in which neutron flux level signals detected by a single neutron detector are separated into a first frequency band region including pulse signals and a second frequency band region including Campbell signals, and these separated signals are amplified under predetermined conditions suitable for each frequency band. A preamplifier for a wide range neutron monitoring system according to this invention is supplied with wide range neutron flux level signals from a neutron detector provided in a reactor and comprises a first amplifier for a high frequency band which includes a low input impedance circuit receiving at the input terminal thereof the wide range neutron flux level signals; a second amplifier for a low frequency band which includes a high input impedance circuit receiving at the input terminal thereof the wide range neutron flux level signals; a third amplifier for amplifying the pulse signals supplied from the output terminal of the first amplifier and for sending the output signals therefrom to a pulse signal counting system; a fourth amplifier for a low frequency band which receives the output signals from the second amplifier and amplifies Campbell signals of a first frequency range, the Campbell signals being included in the output signals from the second amplifier; a fifth amplifier for a high frequency band which receives the output signals from the first amplifier and amplifies Campbell signals of a second frequency range, the Campbell signals being included in the output signals from the first amplifier; and switching means for selectively switching the output signals from the fourth and fifth amplifiers to a Campbelling system.
	abstract	In one embodiment, a fusion reactor includes two internal magnetic coils suspended within an enclosure, a center magnetic coil coaxial with the two internal magnetic coils and located proximate to a midpoint of the enclosure, a plurality of encapsulating magnetic coils coaxial with the internal magnetic coils, and two mirror magnetic coil coaxial with the internal magnetic coils. The fusion reactor further includes one or more electromagnetic wave generators operable to inject a beam of electromagnetic waves into the enclosure.
048620076	summary	BACKGROUND OF THE INVENTION This invention generally relates to casks for holding and transporting radioactive materials, and is specifically concerned with a thermal portection shell for protecting such casks from damage in the event of a fire. Casks for transporting radioactive materials such as the waste products produced by nuclear power plant facilities are known in the prior art. The special purpose of such casks is to ship radioactive wastes as safely as possible. Such casks may be used, for example, to ship high-level vitrified waste cannisters to a waste isolation site, or spent fuel rods to a reprocessing facility. At the present time, relatively few of such transportation casks have been manufactured and used since most of the spent fuel and other wastes generated by nuclear power plants are being stored at the reactor facilities themselves. However, the availability of such on-site storage space is steadily diminishing as an increasing amount of fuel assemblies and other wastes are loaded into the spent-fuel pools of these facilities. Additionally, the U.S. Department of Energy (D.O.E.) has been recently obligated, by the National Waste Policy Act of 1983, to move the spent fuel assemblies from the on-site storage facilities of all nuclear power plants to a federally operated nuclear waste disposal facility starting in 1998. While the transportation casks of the prior art are generally capable of safely transporting wastes such as spent fuel to a final destination, the applicant has observed that there is considerable room for improvement, particularly in the area of fire protection. Nuclear Regulatory Commission (NRC) regulations currently require that Type B casks be capable of withstanding exposure to a fire or other source of infrared radiation which generates a temperature of 1475.degree. F. for at least thirty minutes without significant physical damage, and without the development of excessive internal temperatures and pressures. However, this particular requirement is directly at odds with the requirement that the walls of the casks be capable of conducting, at all times, the heat flux generated by the heat of decay of the radioactive materials contained therein. Accordingly, the problem cannot be solved by merely providing some sort of insulatory sheathing around the casks. While such sheathing might effectively block out excessive heat from a fire of 1475.degree. F. or higher, it would also prevent the heat of decay of the radioactive materials within the casks from dissipating into the ambient atmosphere, thereby causing the build-up of potentially excessive temperatures and pressures within the casks itself. Various attempts have been made to solve the problem by means of an external structure around a cask that acts as a "thermal diode", readily conducting heat from the inside of the cask outwardly, but resisting the conduction of heat from the outside of the cask to its interior. However, all of the prior art attempts to solve the problem require the provision of a relatively delicate and complex structure of thermal cooling fins. Thermal bridges between the cask and the fins are either opened or closed in accordance with the effect of an external source of heat on a material within the fin structure, such as solid lithium and sodium hydroxide, or blocks of aluminum. Unfortunately, such prior art solutions suffer from a variety of shortcomings. For example, the fin structures common to each design are intricate, and require and extensive fabrication effort. Moreover, these fin structures are delicate, and apt to collapse if exposed to a large amount of mechanical shock. Such vulnerability to shock jeopardizes the ability of such prior art thermal shields to operate in a situation where a tractor-trailer transporting the cask is involved in a collision which subjects the cask to a combination of both mechanical shock and fire. Finally, each of these known thermal protection shields are an integral part of the casks which they attempt to protect. Accordingly, none of these shields is easily adaptable for use on a prior art cask which lacks the thermal protection now required by the NRC. Clearly, what is needed is a thermal protection shield which is simple and rugged in construction, and not apt to lose its shielding properties when exposed to a substantial mechanical shock. Ideally, such a shield should further be easily machinable out of common materials so as to minimize the cost of fabrication. Finally, it would be desirable if the thermal shield could easily be used on casks already in existence. SUMMARY OF THE INVENTION Generally, the invention is a thermal protection shell for protecting the exterior walls of a container used for containing heat generating toxic materials, such as radioactive wastes, which fulfills all of the aforementioned criteria. The thermal protection shell comprises a shell of heat conductive material such as aluminum which circumscribes and engages the exterior container walls in intimate, heat-conducting contact under ambient temperature conditions. The thermal coefficient of expansion of the shell is chosen to be substantially larger than the thermal coefficient of expansion of the container walls so that the heat-conducting contact between the shell and the outer wall becomes substantially broken when the shell is exposed to an exterior source of thermal radiation, such as a fire, that raises the temperature of the shell above a selected point. In the preferred embodiment, the value of the thermal coefficient of expansion of the container walls is between about 5 to 7.times.10.sup.6 inches/degrees F. while the value of the thermal coefficient of expansion of the shell is between about 11 to 14.times.10.sup.6 inches/degrees F. The shell made may be detachably mounted onto existing radioactive waste containers. To this end, the shell may be formed from a plurality of sections that are rigidly interconnected by connecting assemblies. In the preferred embodiment, each of these connecting assemblies includes a nut and bolt means for securing mutually adjacent edges of the interconnectable sections together. A second nut means threadedly engaged to the bolt means may be included for both fixing the distance between mutually adjacent edges of the interconnectable sections by removing all slack between the edges, and insuring that the connecting assembly leaves no residual tensile forces in the resulting shell which could interfere with the formation of an insulating gap between the shell and the container in the event of a fire. To further insure the connecting assembly will not interfere with the formation of the gap, the nut and bolt means are preferably also formed from a material, having a substantially higher thermal coefficient expansion than the material forming the walls of the container. It should be noted that the adjustability provided by such nut and bolt connecting assemblies advantageously allows a particular shell which has been fabricated for a particular model of cask to fit any such cask despite the dimensional differences caused by machine tolerances. In the preferred embodiment, the shell has sufficient mass and a melting temperature close enough to the temperature causing gap formation so that the ablation of the shell also serves to obstruct the transmission of heat from the source of thermal radiation to the container walls. In the preferred embodiment, the walls of the container are formed from steel, and the shell is approximately one-half of an inch thick, and formed from either an aluminum, aluminum alloy, magnesium, or a magnesium alloy.
	abstract	Example embodiments are directed to jet pump inspection apparatuses including an encoded probe driver coupled to a guide funnel. In an example embodiment, the encoded probe driver and the guide funnel are located on the body of the apparatus in order to decrease the time required for inspection, thereby limiting exposure to personnel of radioactive contamination.
	claims	1. A control rod blade for a boiling water reactor, comprising a plurality of channels each of a predetermined diameter and oriented about a longitudinal axis, said channels being arranged to receive an absorber material, a free edge portion with a recess, which includes outlets for said channels, and a cover element having a cover portion, said cover element being sealingly attached to said free edge portion to seal said recess, said cover portion is positioned outside of said recess and forms an external end surface of said control rod blade in a mounted state, a profile element having a thickness less than the diameter of the channels, wherein said thickness is measured along an axis of said profile element coaxial with said longitudinal axis, said profile element being disposed against a bottom surface defined by the recess; said profile element covers the outlets of said channels and wherein said profile element and said cover element engage one another and cooperate to define at least one passage there between. 2. A control rod blade according to claim 1, wherein the profile element has a width which substantially corresponds to a width defined by the bottom surface. 3. A control rod blade according to claim 1, wherein the profile element comprises a substantially plane surface, which is arranged to be applied against a corresponding substantially plane bottom surface. 4. A control rod blade according to claim 1, wherein the profile element comprises at least one curved side portion, which has an extension projecting outwardly from a substantially plane surface. 5. A control rod blade according to claim 1, wherein the profile element has a thickness of about 0.2-0.5 mm. 6. A control rod blade according to claim 1, wherein the profile element has a continuous extension along a whole length of the recess. 7. A control rod blade according to claim 1, wherein the profile element is manufactured of a metal material. 8. A control rod blade according to claim 1, wherein the cover element comprises a surface, which is arranged to abut a surface of the profile element when the cover element is applied in the recess. 9. A control rod blade according to claim 8, wherein the contact surfaces of the profile element and the cover element are substantially plane. 10. A control rod blade according to claim 1, wherein the cover element comprises support portion, which has a width, which is less than a width defined by the recess. 11. A control rod blade according to claim 1, wherein the recess comprises a groove which, after that the profile element has been applied in the recess, is arranged to form a passage, which extends between adjacent channels under the profile element. 12. A control rod blade according to claim 1, wherein the cover element is arranged to be attached at the edge portion of the control rod blade by means of two longitudinal weld joints. 13. A control rod blade according to claim 1, wherein the absorber material is powdered. 14. A control rod blade according to claim 13, wherein the absorber material comprises boron carbide.
	summary
039322157	summary	The invention relates to a vertical control rod for shutting off nuclear reactors rapidly and safely. More particularly, the control rod is of the type that is removable from the reactor in an upward direction, and is formed of a flexible absorber guided in a tube, an acceleration spring, a drive mechanism for tensioning the spring, and a holding device for the spring. Flexible absorbers are employed to avoid dangerous jamming of the absorbers in the guide therefor, particularly in the case of very long control rods and are known, for example, from German Pat. Nos. 1,092,572 and 1,100,831. In these and other shut-off devices, the flexible absorbers are suspended from a holding device above the reactor core during the operation of the reactor and are dropped into the reactor core, accelerated by gravity in order to shut off the reactor. Shut-off devices of this kind are very reliable, but the shut-off speed thereof and the forces available for overcoming resistances are limited. Any additional force acting upon the suspended, flexible absorbers from above and exceeding the force of gravity, however, increases the danger of jamming in the guide. It is accordingly an object of the present invention to provide a vertical control rod for rapidly and safely shutting off a nuclear reactor which is removable from the reactor in an upward direction, and has a force available for effecting shut-off which exceeds the force of gravity, yet avoids jamming of the absorber in the guide therefor, even when, in the event of damage, deformation of the guide tube might result. With the foregoing and other objects in view, there is provided, in accordance with the invention, vertical control rod for rapidly and safely shutting off a nuclear reactor, the rod being removable from a reactor in upward direction and comprising an elongated tube, a flexible absorber guided in the tube, spring means operatively connected to the absorber for accelerating vertical displacement thereof, a drive mechanism for tensioning the spring, means for holding the spring, a traction device suspending the absorber below the core of the reactor during operation of the reactor and means for disposing the accelerating spring means and the drive mechanism above the reactor core. The flexible absorber is thereby stressed only in tension and jamming thereof in the guide is thereby avoided. In accordance with a further feature of the invention, the absorber is disconnectible from the traction device therefor, and the drive mechanism above the reactor core, while being simultaneously lockable into its shut-off position. Thus, the entire control rod, except for the absorber part, is exchangeable without impairing the shut-off action. In accordance with another feature of the invention, the absorber is moveable into shut-off position by means of the drive mechanism for tensioning the acceleration spring. While sacrificing shut-off speed, a result of this construction is that the absorber is moved, with greater reliability, into the shut-off position if the acceleration spring should fail. It is also advantageous so to control the drive mechanism that it follows the absorber into the shut-off position immediately after the shut-off is effected. In accordance with yet another feature of the invention, the traction device has a hollow construction and contains a wire rope secured at both ends thereof. This construction affords an additional safety precaution and assures a strong connection between the absorber and the drive mechanism even if the traction device itself should break. In accordance with an additional feature of the invention, the absorber is formed of several members that are movable with respect to each other and are guidable in a guide tube which is traversible by a coolant flow, the lowermost absorber being disposed, during the operation of the reactor in a lower end section of the guide tube, substantially sealing the spacing between the absorber and the inner surface of the guide tube from the outside. This construction has the advantage that, in the event of shut-off, the absorber is sufficiently cooled by a flow of coolant between the absorber and the inner surface of the guide tube but that, during the operation of the reactor, no appreciable or substantial quantities of coolant can flow through the guide tube which would contribute only little to the cooling of the reactor core.
	description	This application claims priority to German application No. 10 2013 214 393.1 DE filed Jul. 23, 2013, the entire content of which is hereby incorporated herein by reference. The invention relates to an X-ray radiography system for differential phase contrast imaging of an object under investigation by means of phase-stepping, with an X-ray emitter for generating a beam path of quasi-coherent X-ray radiation, an X-ray image detector with pixels arranged in a matrix, and a diffraction or phase grating, in which the X-ray emitter has an X-ray tube with a cathode and an anode. Grating-based phase contrast imaging (PCI) is a relatively new imaging method by which, instead of an X-ray absorption image, measurement data is recorded which makes it possible to obtain in parallel both an absorption-based X-ray image, a differential phase image and a dark field image. From the data in the two additional images, further information can be obtained which can, for example, be used in a clinical diagnosis. A requirement for grating-based phase contrast imaging is local coherence of the X-ray radiation, at least in one direction. A further technological bottleneck is that the interference pattern which is to be recorded normally has a smaller periodic resolution than that which can be resolved by standard X-ray image detectors. The standard solution to this is to use a so-called phase-stepping technique, which will be explained in more detail below by reference to FIG. 2. With this, an analyzer grating G2 with the periodicity of the undisturbed interference pattern is set up in front of the X-ray image detector, and is displaced relative to the object and the X-ray image detector, where the magnitude of the displacement is less than one period of the grating. From three or more one-shot recordings, the interference pattern can then be reconstructed. For differential phase contrast imaging (PCI), it is usual to insert three gratings in the beam path from the source of the X-ray beam. The article “Phase retrieval and differential phase-contrast imaging with low-brilliance X-ray sources” by Franz Pfeiffer et al., which appeared in Nature Physics 2 (2006), pages 258 to 261, describes an example of this type of PCI, which will be explained in more detail below. The wave nature of particles such as X-ray quanta permits phenomena like diffraction and reflection to be described with the help of the complex refractive indexn=1−δ+iβ. Here, the imaginary component β describes the absorption, on which is based today's clinical X-ray imaging, such as for example computed tomography, angiography, radiography, fluoroscopy or mammography, and the real component δ describes the phase shift which is used in differential phase imaging. From DE 10 2010 018 715 A1, an X-ray radiography system is known in which, for the purpose of high-quality X-ray imaging, for the purpose of phase contrast imaging of an object under investigation use is made of an X-ray radiography system which has at least one X-ray emitter with a plurality of field emission X-ray sources for emitting coherent X-ray radiation, an X-ray image detector, a diffraction grating G1 arranged between the object under investigation and the X-ray image detector, and another grating G2 which is arranged between the diffraction grating G1 and the X-ray image detector. An X-ray recording system, with which differential phase contrast imaging of the type mentioned in the introduction can be carried out, is known for example from U.S. Pat. No. 7,500,784 B2, and this will be explained by reference to FIG. 1. FIG. 1 shows the typical main features of an X-ray radiography system for an intervention suite with a C-arm 2, held by a stand 1 in the form of a six-axis industrial or articulated arm robot, attached to the ends of which are an X-ray radiation source, for example an X-ray emitter 3 with X-ray tubes and collimator, and an X-ray image detector 4 as the image recording unit. Using, for example, the articulated arm robot known from U.S. Pat. No. 7,500,784 B2, which preferably has six axes of rotation and hence six degrees of freedom, the C-arm 2 can be moved to any desired position in space, for example by being turned about a center of rotation between the X-ray emitter 3 and the X-ray image detector 4. The inventive angiographic X-ray system 1 to 4 can, in particular, be rotated about centers of rotation and axes of rotation in the plane of the C-arm of the X-ray image detector 4, preferably about axes of rotation which pass through the mid-point of the X-ray image detector 4 and the mid-point of the X-ray image detector 4. The familiar articulated arm robot has a basic frame which, for example, has a fixed mounting on the floor. To this is attached a carousel that can rotate about a first axis of rotation. Attached to the carousel so that it can pivot about a second axis of rotation is a robot swing arm, to which is affixed a robot arm which can rotate about a third axis of rotation. At the end of the robot arm is attached a robot hand which can rotate about a fourth axis of rotation. The robot hand has a fixing element for the C-arm 2, which can pivot about a fifth axis of rotation and can rotate about a sixth axis of rotation oriented at right angles to the fifth. The implementation of the X-ray diagnostic facility is not dependent on an industrial robot. It is also possible to make use of the usual C-arm or mammography devices. The X-ray image detector 4 can be a rectangular or square flat semiconductor detector, preferably manufactured from amorphous silicon (a-Si) or selenium (a-Se). However, use could also be made of integrating and possibly counting CMOS detectors. In the beam path from the X-ray emitter 3 there is, on a table plate 5 of a patient positioning table, a patient 6 who is to be investigated as the object under investigation. Connected to the X-ray diagnostic facility is a system control unit 7 with an imaging system 8, which receives and processes the image signals from the X-ray image detector 4 (examples of the operating elements are not shown). The X-ray images can then be inspected on displays on a suspended monitor 9. The suspended monitor 9 can be held by means of a ceiling-mounted, longitudinally movable, pivotable, rotatable and height-adjustable carrier system 10 with a cross-arm and lowerable carrier arm. Instead of the X-ray system illustrated by way of example in FIG. 1, with its stand 1 in the form of a six-axis industrial or articulated arm robot, the angiographic X-ray system could also have a normal ceiling or floor mounted holder for the C-arm 2. Instead of the C-arm 2 illustrated by way of example, the angiographic X-ray system could also have separate ceiling and/or floor mounted holders for the X-ray emitter 3 and the X-ray image detector 4, which are, for example, rigidly coupled to each other electronically. In today's known arrangements for clinical phase contrast imaging, use is made of conventional X-ray tubes, X-ray image detectors which are today available, such as for example those described by Martin Spahn in “Flachbilddetektoren in der Röntgendiagnostik” [Planar image detectors in X-ray diagnostics], Der Radiologe, Volume 43 (5-2003), pages 340 to 350, and three gratings, G0, G1 and G2, as will be explained in more detail by reference to FIG. 2, which shows schematically a structure for a Talbot-Lau interferometer for differential phase contrast imaging, with extended tube focus, gratings G0, G1 and G2 and a pixelated X-ray image detector. For the purpose of producing coherent radiation, the X-ray beams 12 emerging from a tube focus 11 of the non-coherent X-ray emitter 3 pass through an absorption grating 13 (G0) which effects local coherence, and through an object under investigation 14, for example the patient 6. The object under investigation 14 deflects the wave front of the X-ray beam 12 by phase shifting in a way that is made clear by the normal 15 to the wave front when there is no phase shift, i.e. with no object, and the normal 16 to the wave front with phase shifting. Following this, the phase-shifted wave front passes through a diffraction or phase grating 17 (G1), which has a grating constant matched to the mean energy of the X-ray spectrum, for the purpose of producing interference lines (Talbot effect), and then in turn through an absorptive analyzer grating 18 (G2) for reading out the interference pattern produced. The grating constant of the analyzer grating 18 is matched to that of the phase grating 17 and to the remaining geometry of the arrangement. The analyzer grating 18 is, for example, arranged at the first or nth Talbot interval. By so-called “phase-stepping”, described below, together with the analyzer grating 18 (G2), it is possible to detect relevant items of data from the interference pattern, using the X-ray image detector 4. If the tube focus 11 of the X-ray radiation source is sufficiently small, and the radiation power generated is nevertheless sufficiently large, it may be possible to forgo the first grating G0, the absorption grating 13, as would be the case for example if a plurality of field emission X-ray sources are provided as the X-ray emitter 3, as is known from DE 10 2010 018 715 A1 which has been described. For each pixel of the X-ray image detector 4, the image data is now determined in that, by the phase stepping 19, which is indicated by an arrow, the analyzer grating 18 (G2) is moved in several steps (k=1, K, where for example K=3 to 8) by an appropriate fraction of the grating constant, perpendicularly to the direction of radiation of the X-ray beams 12 and laterally relative to the arrangement of the grating structure, and the signal Sk which arises for this configuration during the recording is measured in the pixel of the X-ray image detector 4, and by this means the interference pattern which arose is sampled. For each pixel, the parameters of a function (e.g. a sine function) which defines the modulation is then determined by a suitable fitting method, by a matching or compensation method applied to the signals Sk measured in this way. The visibility, i.e. the normalized difference between the maximal and minimal signals (or more precisely: the amplitude normalized to the mean signal), is here a measure for characterizing the quality of a Talbot-Lau interferometer. It is defined as the contrast of the sampled modulation V = I max - I min I max + I min = A I _ . Further, in this equation A represents the amplitude and Ī the mean intensity. The visibility can take on values between zero and one, because all the variables are positive and Imax>Imin. In a real interferometer it is also the case that Imin>0, so that the value of V meaningfully extends across the whole range. Minimal intensities greater than zero, and all non-ideal characteristics and defects of the interferometer, lead to a reduction in the visibility. A third item of data which can be defined by means of the visibility and is generated by this type of measurement is called the dark field. The dark field gives the ratio of the visibilities of measurements with an object and those without an object. D = V obj V ref = A obj · I _ ref A ref · I _ obj . From a comparison of certain derived quantities from the functions fitted for each pixel, once with and once without an object (or patient), it is then possible to produce three different images: (i) an absorption image, (ii) a differential phase contrast image (DPC) and (iii) a dark-field image. In the case of the known PCI imaging methods, use is currently made of either a microfocus X-ray tube, which of itself meets the required coherence conditions, or alternatively an absorption grating with the designation G0, which splits up the array of X-ray beams, output from the anode of the X-ray tube, into lines of X-rays. Each of these lines taken on its own satisfies the coherence condition and is so positioned by its spacing from its neighboring lines that in the plane of the detector the interference images overlay each other constructively in conformity with the method according to Lau. The disadvantage of this method is that a large proportion of the X-ray radiation generated is absorbed in the absorption grating G0, because the ratio of the opening to absorptive material is ≦1. The invention is based on the object of designing an X-ray radiography system of the type mentioned in the introduction such that it permits particularly simple phase contrast imaging, with real time capability, with a high resolution and alternative local coherence of the X-ray radiation. For an X-ray radiography system of the type mentioned in the introduction, this object is achieved in accordance with the invention by the features specified in claim 1. Advantageous developments are specified in the dependent claims. For an X-ray radiography system, the objective is inventively achieved by constructing the X-ray tube in such a way that an electron ray beam originating from the cathode is associated with focusing electronics which produce, from electrons incident on an anode, at least one linear-shaped electron fan beam. The effect of this is that the absorption grating G0 is no longer necessary, so that almost all of the X-ray radiation which is produced can be used for the imaging. It is advantageous that the conventional phase stepping can be replaced by a magnetic deflection facility, associated with the linear-shaped electron fan beam, which varies the point of incidence of the electrons on the anode so that a mechanical movement of the analyzer grating G2 is no longer necessary. It has proved to be advantageous if the focusing electronics produce at least one linear set of electron fan beams from electrons incident on an anode. In accordance with the invention, the spacing between the lines of the linear set of electron fan beams can be dimensioned so that it satisfies the Lau condition, the constructive overlaying of the interference patterns at the site of the image plane. It is advantageous if the focusing electronics are constructed in such a way that the spacing between the lines in the linear set is changed by actuation of the focusing electronics. In accordance with the invention, it is possible to carry out an equivalent phase stepping if the linear-shaped electron fan beams are moved perpendicularly to the direction of the lines. The X-ray tube of the X-ray emitter 3, equipped in accordance with the invention, is modified in such a way that electron ray beams 21 emerging from the cathode 20 are influenced by focusing electronics 22 in such a way as to form a first linear set 23 of linear-shaped electron fan beams. These can be yet further refined into a second linear set 24 of linear-shaped electron fan beams. A magnetic deflection device 25 can vary the points of incidence of the second linear set 24 on an anode 26, for example a stationary anode. Within the X-ray tube of the X-ray emitter 3, the conventional focusing of the electrons onto the anode 26 is modified. The focusing electronics 22, which can consist of a lens and grating system, are constructed in such a way that the electrons are focused into lines each with a size, for example, of 4 μm×450 μm, so that the sets of lines 23 and 24 are formed. The spacing between the lines satisfies the Lau condition, the constructive overlaying of the interference patterns at the site of the image plane. Optionally, it is possible to alter the spacing by actuation of the focusing electronics 22. Furthermore, the linear set 24 can be deflected by the magnetic field of the magnetic deflection device 25, as described for example in U.S. Pat. No. 8,284,894 B2. If the linear set 24 is moved perpendicularly to the direction of the lines, an equivalent phase stepping can be carried out by this means. In the extreme case, the focus could consist of a single line (cf. DE 102 45 676 B4). The conventional phase stepping 19 can then be replaced by a suitable magnetic deflection of the focus. It would also be possible to build up a chessboard pattern by electron focusing, thus producing coherence in two directions. It would then be possible to effect magnetic deflection in both directions or a “diagonal” movement. The inventive construction of the X-ray tube of the X-ray emitter 3 enables almost all the X-ray radiation generated to be utilized for imaging purposes. Furthermore, the absorption grating G0 is no longer necessary. A mechanical movement of the analyzer grating 18 (G2) is also no longer required (phase stepping). It is sufficient to vary the magnetic field by means of the magnetic deflection device 25. Using the inventive arrangement, one obtains an alternative local coherence of the X-ray radiation. Furthermore, there is a simplification of the PCI imaging method due to the replacement of the normal phase stepping 19 by the electrostatic and magnetic deflection of the electron ray beams 21 or the linear-shaped electron fan beams 23 and/or 24, as applicable.
	abstract	A pressure vessel penetration sidewall adjacent a tube installed in the penetration by a clearance fit is inspected by passing an eddy current probe having a pair of circumferential coils through the tube. Eddy currents are induced in the pressure vessel as the probe passes through the penetration tube and degradation of the pressure vessel adjacent the clearance is determined based upon the eddy currents induced in the pressure vessel by the probe.
062326112	abstract	A radiographic intensifying screen having at least a fluorescent layer and a protective layer on a support, wherein the protective layer has a multi-layer structure comprising at least one layer of an organic macromolecule film and a film-forming resin layer provided on the surface of the organic macromolecule film at least on the side which is not in contact with the fluorescent layer, and the resin of the film-forming resin layer is different from the resin of the organic macromolecule film.
	claims	1. An EUV collector configured to transfer an emission of at least one EUV radiation source to a main intensity spot, the EUV collector comprising:a collector subunit comprising first and second grazing incidence mirrors arranged one behind the other, the first and second grazing incidence mirrors being configured to transfer EUV radiation from the radiation source to a transfer intensity spot; andan ellipsoid mirror having an ellipsoidal mirror surface configured so that, during use of the EUV collector, the EUV radiation impinges on the ellipsoidal mirror surface at an angle of incidence below a critical grazing incidence angle,wherein:the EUV collector has no more than one collector subunit comprising any grazing incidence mirrors in a path of an EUV partial beam between the EUV radiation source and the transfer intensity spot;EUV partial beams are emitted by the EUV radiation source so that each EUV partial beam is reflected by no more than one single collector subunit comprising at least one grazing incidence mirror;at least some of the EUV partial beams emitted by the EUV radiation source are reflected only by the collector subunit comprising the first and second grazing incidence mirrors;the collector subunit is configured to collect a forward emission portion of the emission of the at least one EUV radiation source which is emitted toward the main intensity spot; andthe ellipsoid mirror is configured to collect a backward emission portion of the emission of the at least one EUV radiation source which is emitted away from the main intensity spot. 2. The collector of claim 1, wherein the collector comprises exactly one collector subunit which comprises at least one grazing incidence mirror. 3. The collector of claim 1, wherein the collector comprises exactly two collector subunits which comprise at least one grazing incidence mirror. 4. The collector of claim 1, wherein the collector subunit comprising the first and second grazing incidence mirrors is configured to transfer the EUV radiation from the radiation source to an intermediate intensity spot. 5. The collector of claim 4, comprising a plurality of collector subunits is arrangeable around the EUV radiation source, each of the plurality of collector subunits comprises at least one grazing incidence mirror. 6. The collector of claim 1, wherein the collector comprises exactly one collector subunit which comprises at least one grazing incidence mirror and an ellipsoid mirror in the path of the EUV partial beam emitted by the radiation source between the EUV radiation source and the main intensity spot. 7. The collector of claim 1, wherein the collector is arranged so that EUV radiation reflected by the ellipsoid mirror is guided directly to the main intensity spot via a through-opening in the collector subunit comprising the first and second grazing incidence mirrors. 8. The collector of claim 1, wherein the collector is arranged so that EUV radiation reflected by the ellipsoid mirror is guided directly to the main intensity spot via a spatial region which is around the collector subunit comprising the first and second grazing incidence mirrors. 9. An illumination optics, comprising:a collector according to claim 1 configured to illuminate an object field which is imageable by an imaging optics. 10. An illumination system, comprising:at least one EUV radiation source; andan illumination optics comprising a collector according to claim 1 configured to illuminate an object field which is imageable by an imaging optics. 11. The illumination system of claim 10, comprising a plurality of EUV radiation source. 12. An apparatus, comprising:an illumination system comprising a collector according to claim 1;an imaging optics configured to image an object field into an image field,wherein the illumination optics is configured to illuminate the object field of the imaging optics, and the apparatus is a projection exposure apparatus. 13. The apparatus of claim 12, further comprising at least one EUV radiation source. 14. The apparatus of claim 12, wherein the collector comprises exactly one collector subunit which comprises at least one grazing incidence mirror. 15. A method of using a projection exposure apparatus comprising an illumination system and an imaging optics, the method comprising:using the illumination system to illuminate a reticle with EUV radiation; andusing the imaging optics to project at least a portion of the reticle onto a material which is sensitive to the EUV radiation,wherein the illumination system comprises a collector according to claim 1. 16. A collector, comprising:a plurality of ellipsoid mirrors comprising a first ellipsoid mirror and a second ellipsoid mirror,wherein:the collector is configured to transfer emission of EUV radiation from a plurality of EUV radiation sources to a single main intensity spot; andthe radiation sources are disposed at both ellipse focus points of at least one of the ellipsoid mirrors. 17. The collector of claim 16, further comprising:a collector subunit comprising first and second grazing incidence mirrors arranged one behind the other, the first and second grazing incidence mirrors being configured to transfer EUV radiation from the radiation source to a transfer intensity spot,wherein:the first ellipsoidal mirror surface is configured so that, during use of the EUV collector, the EUV radiation impinges on the first ellipsoidal mirror surface at an angle of incidence below a critical grazing incidence angle,EUV partial beams are emitted by a first EUV radiation source of the plurality of EUV radiation sources so that each EUV partial beam is reflected by no more than one single collector subunit comprising at least one grazing incidence mirror; andat least some of the EUV partial beams emitted by the first EUV radiation source are reflected only by the collector subunit comprising the first and second grazing incidence mirrors. 18. The collector of claim 17, wherein the collector comprises exactly one collector subunit which comprises at least one grazing incidence mirror. 19. The collector of claim 17, wherein the collector comprises exactly two collector subunits which comprise at least one grazing incidence mirror.
060758406	abstract	Air cross grids, for absorbing scattered secondary radiation and improving radiation imaging in general radiography and in mammography, are provided with a large plurality of open air passages extending through each grid panel. These passages are defined by two large pluralities of substantially parallel partition walls, respectively extending transverse to each other. Each grid panel is made by laminating a plurality of thin metal foil sheets photo-etched to create through openings defined by partition segments. The etched sheets are aligned and bonded to form the laminated grid panel, which is moved edgewise during the radiation exposure to pass primary radiation through the air passages while absorbing scattered secondary radiation arriving along slanted paths.
052672875	description	DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 is a plan view of a nuclear fuel assembly 10, in which only the upper end fitting 12 is visible. The upper portion of the fuel assembly 10 is shown in elevation in FIG. 2, immediately below, and spaced from, a horizontally oriented upper core support plate 100. As would be found in a nuclear reactor core. In the reactor core, a plurality of the vertically oriented fuel assemblies are positioned in side by side relationship. After all the fuel assemblies are in place, the upper core support plate 100 is lowered on to all of the assemblies, substantially simultaneously. Each fuel assembly has a multiplicity of nuclear fuel rods 16 which are connected to the base 14 of the upper end fitting 12. The base includes a plurality of holes 18 by which reactor coolant passes into and through the upper end fitting 12, from the flow channels between the fuel rod 16. The upper end fitting in one conventional arrangement as shown in FIGS. 1 and 2, includes four spring packs or members 20, 22, 24, and 26, each of which extends along one respective side of the substantially square end fitting perimeter. In the illustrated embodiment, spring member 20 is situated between corners 50 and 44, spring member 22 between corners 44 and 46, spring member 24 between corners 46 and 48, and spring member 26 between corners 48 and 50. The spring members project upwardly to define the upper limit of the fuel assembly. Each spring member, such as 24 shown in FIG. 2, includes a rigidly supported portion 28, and a free end 30, between which an active portion or surface 32 is adapted to contact and bear against the support plate 102. In other words, as the support plate 100 is lowered onto the fuel assemblies, contact is made between the surface 102 of the plate 100 and the portion 32 of each spring such as 24, whereby the spring members are simultaneously loaded. The support plate 100 is then rigidly secured to the reactor vessel (not shown), such that all the fuel assemblies are resiliently supported during reactor operation. In the embodiment shown in FIGS. 1 and 2, each spring member 24 is cantilevered. Moreover, it is preferred that each spring member 24 include a plurality of nested, cantilevered spring elements including elements 34 and 36. Spring elements 34 and 36 are rigidly connected at one end to spring portion 28 and these are attached near the corner 46 of the end fitting. At their free ends, elements 34 and 36 have openings 38, 40 through which the free end 30 of the main spring element passes substantially vertically. It should be appreciated that as the plate 100 is lowered onto bearing surface 32, and during flexure of the spring member 24 in operation, the interaction between surfaces 32 and 102, and the mutual points of contact among the spring element such as shown at 42, have a sliding e.g., horizontal, component. This friction not only affects the spring rate, but, particularly where vibratory forces on the fuel assembly are manifested at the spring contact surfaces, can give rise to excessive wear and corrosion. This can pose a problem even when inherently corrosion resistant materials are used for the springs and core support plate. For example, it is typical that the elements of spring member 24 are made from Inconel, e.g., Inconel 718, whereas the core support plate 100 is made from stainless steel. Nevertheless, according to the present invention, there is achieved an enhanced lubricity between each spring member such as 24, against the core support plate 100, and preferably, between the spring elements such as 34, 36, and 28, of a given spring member such as 24. This enhanced lubricity is accomplished by coating at least the active, bearing surface 32 of the spring member, with a smooth metallic material. Metal nitrides, particularly ZrN and TiN, are especially effective. Tests on representative samples for the interaction of these nitride coatings on Inconel 718 show significant enhancement of desirable characteristics. TABLE 1 ______________________________________ Properties of Nitride Coated Inconel 718 Property ZrN TiN ______________________________________ Microhardness 2,895 2,575 (Kg/mm.sup.2) (50 g load) Surface Roughness 0.12 0.38 (rms) (um) Coefficient of Friction 0.020 0.035 ______________________________________ The wear rate of Inconel 718 is reduced by a factor of six, and more importantly for the present invention, the frictional force between the coated spring and the stainless steel core plate, is reduced by about a factor of eight. These data are base on coating Inconel 718 samples using a cathodic vacuum arc plasma deposition process as described, for example, in the article "Cathodic Arc Deposition Advances in Coating Technology", P.C. Johnson, Research and Development, February, 1987. It should be appreciated, however, that other coating processes may be employed to achieve the advantages within the scope of the present invention. Although a given process may be more convenient or cost effective than another, the novelty of coating the spring members on the end fittings of nuclear fuel assemblies, to enhance lubricity, is not dependent on the particular process selected. Other coating materials which can provide significant improvement relative to the current practice of using uncoated spring members, include the metal nitrides CrN, HfN, TiAlVN, TaN, and TICN. In addition, other suitable coatings include Cr, TiC, CrC, ZrC, and NiTaB. Although, as a minimum, the bearing surface such as 32 as shown in FIG. 2, is enhanced by means of the coating, alternatively the entire external surface of the spring member 24, or of each of the spring elements constituting the spring member, can be coated. This enhances lubricity at the contact surfaces, and reduces the risk of corrosion at any other spring member surface. FIG. 3 illustrates another upper end fitting embodiment 200, in which two leaf spring members 202, 204, perform the same function as the four cantilever spring members of the embodiment shown in FIGS. 1 and 2. The leaf spring member 202 is rigidly connected near corners 48 and 44, with the apex 206 substantially above corner 46. Similarly, spring member 204 is rigidly connected at its ends near corners 44 and 48, with the apex 214 substantially above the corner (not shown) opposite corner 46. The leaf spring member such as 202,, is preferably formed from two nested leaf spring elements 208, and 210. The bearing surface 206 at the apex is adapted to contact the upper core support plate, deflect when loaded, and experience flexure during operation, in a manner analogous to that described above with respect to spring member 24. The leaf spring embodiment shown in FIG. 3, has a greater contact surface between the respective spring elements 208, and 210, and therefore would benefit significantly from the enhanced lubricity on the full exterior surfaces of each spring element, in accordance with the present invention.
	summary
	claims	1. A grid module of a scattered-radiation grid comprising a number of grid modules with webs arranged next to one another, a height of at least one of the webs disposed on an edge side in a respective one of the grid modules is relatively lower than a height of webs arranged further inwards in the respective grid module. 2. The grid module of claim 1, wherein a height of at least one further of the webs is embodied relatively increasing in height inwards in stages from an at least one edge side relatively lower web. 3. A detector of a CT system comprising a modular construction scattered-radiation grid, the scattered-radiation grid including the grid module of claim 1. 4. A CT system comprising a detector including a modular construction scattered-radiation grid, the scattered-radiation grid including grid modules of claim 1. 5. A scattered-radiation grid for an x-ray detector of a CT system including a plurality of detector elements disposed in rows and columns over its surface, comprising:at least two grid modules arranged next to one another, each of the at least two grid modules possessing a number of webs disposed next to one another with irradiation zones between them, and at least one edge-side web of at least one of the grid modules being adjacent to at least one other web disposed running in parallel next to one another on the edge side of another of the grid modules in absence of an irradiation zone arranged between them, webs running adjacent to each other including a relatively lower height than other webs. 6. The scattered-radiation grid of claim 5, wherein relatively lower height of the webs running adjacent to one another is dimensioned so that the proportion of scattered radiation additionally absorbed by the webs running adjacent to one another is compensated for by the reduced relative height of the webs. 7. The scattered-radiation grid of claim 5, wherein at least one further of the webs disposed further inwards in relation to the at least one grid module has a height which lies between the height of the webs disposed even further inwards and a web disposed further outwards. 8. The scattered-radiation grid of claim 7, wherein the relatively lower height of the webs arranged further inwards is dimensioned such that the proportion of scattered radiation additionally absorbed by the webs running adjacent to one another is compensated for by the reduced height of the webs. 9. The scattered-radiation grid of claim 5, wherein webs disposed exclusively in parallel to one another are provided. 10. The scattered-radiation grid of claim 5, wherein webs crossing each other at right angles are provided. 11. The scattered-radiation grid of claim 10, wherein the height of at least one web reduces in stages at its ends. 12. The scattered-radiation grid of claim 6, wherein at least one further of the webs disposed further inwards in relation to the at least one grid module has a height which lies between the height of the webs disposed even further inwards and a web disposed further outwards. 13. The scattered-radiation grid of claim 12, wherein the relatively lower height of the webs arranged further inwards is dimensioned such that the proportion of scattered radiation additionally absorbed by the webs running adjacent to one another is compensated for by the reduced height of the webs. 14. A detector of a CT system comprising the scattered-radiation grid of claim 5. 15. A CT system comprising a detector including the scattered-radiation grid of claim 5.
	summary
041982711	claims	1. In a liquid metal cooled nuclear reactor construction comprising a primary vessel housed with a concrete containment vault, the primary vessel being suspended from the roof structure of the containment vault, a reactor core submerged in a pool of coolant in the primary vessel, the reactor core being supported from the wall of the base region of the primary vessel, a leak jacket spaced from and surrounding the primary vessel, control rods supported from the roof structure of the vault and insertable in the core, the improvement comprising secondary core support means, nonloaded during normal operation of the reactor said means being located in the interspace between the primary vessel and the leak jacket for limiting the extent of downward deflection of the base region of the primary vessel. 2. A liquid metal cooled nuclear reactor construction according to claim 1 wherein said means for limiting the extend of downward deflection of the base region of the primary vessel comprises an abutment member attached to the leak jacket in manner providing a clearance between the member and the primary vessel. 3. A liquid metal cooled nuclear reactor construction according to claim 2 wherein the leak jacket is suspended from jacking means of the roof structure by tie members, and the clearance between the abutment member and the primary vessel is adjustable by the jacking means. 4. A liquid metal cooled nuclear reactor construction according to claim 3 wherein the reactor core is supported from the wall of the base region of the primary vessel by means of a strongback and diagrid, combination which has means for limiting downward displacement of the diagrid relative to the strongback.
06297419&	claims	1. A process for the treatment of zirconium based metal waste, the process including the steps of converting at least some of said zirconium based metal to an oxide thereafter forming a mixture of said oxide with a sol or a solution of a suitable gel forming material producing a green body from the mixture, and sintering said green body into a sintered body. 2. A process according to claim 1 wherein said material hardens by hydrolysis. 3. A process according to claim 2 wherein said gel forming material comprises a metal salt. 4. A process according to claim 3 wherein said salt is an alkoxide. 5. A process according to claim 4 wherein said alkoxide is aluminum secondary butoxide or aluminum isopoxide. 6. A process according to claim 1 wherein the gel forming material is treated with a modifying agent. 7. A process according to claim 6 wherein said modifying agent is an alkanolamine. 8. A process according to claim 7 wherein said alkanolamine is triethanolamine. 9. A process according to claim 1 wherein the zirconium oxide is mixed with a hydrolyzed zirconium salt or other metal salt. 10. A process according to claim 1 wherein the green body is dried prior to sintering. 11. A process according to claim 1 wherein the green body is formed in a desired shape by casting in a mold or by extrusion. 12. A process according to claim 1 wherein additional metal oxides are present in said green body to stabilize the chemical phases of said sintered body. 13. A process according to claim 1 wherein the sintering temperature is in the region of 1400.degree. C. 14. A process according to claim 1 wherein said zirconium based metal is brought into solution by chemical or by electrochemical dissolution. 15. A process according to claim 14 wherein said oxide includes at least some zirconium oxide formed directly as a result of said dissolution. 16. A process according to claim 14 wherein said oxide includes at least some zirconium oxide formed by conversion from said zirconium based metal in solution. 17. A process according to claim 14 wherein the metal solution is in the form of a nitrate. 18. A process according to claim 17 wherein the nitrate is thermally treated to decompose the nitrate to the oxide. 19. A process according to claim 14 wherein the zirconium based metal is brought into solution by means of dissolution of a fluoride. 20. A process according to claim 1 wherein the zirconium based metal waste is derived from the cladding of a nuclear fuel rod. 21. A process according to claim 20 wherein the zirconium based metal is derived from whole fuel rods comprising irradiated fuel. 22. A process according to claim 21 wherein said zirconium based metal is brought into solution by chemical or by electrochemical dissolution, said oxide includes at least some zirconium oxide formed by conversion from said zirconium based metal in solution, and fuel constituents of the solution are separated prior to said conversion of the solution to the oxide. 23. A process according to claim 22 wherein at least some fission product constituents remain with the zirconium solution and the resulting oxide. 24. A process according to claim 1 wherein the sintered body is hot pressed to densify it. 25. The process for the treatment of zirconium based metal waste of claim 1 further comprising the step of storing said sintered body. 26. A process for the treatment of zirconium based metal waste, the process including the steps of converting at least some of said zirconium based metal to an oxide, thereafter forming a mixture of said oxide with a suitable material which gels and harden by heat, producing a green body from the mixture, and sintering said green body. 27. A process according to claim 26 wherein said material is selected from the group consisting of zirconium acetate, zirconium acetate combined with citric acid, zirconium nitrate combined with citric acid and zirconium acrylamide. 28. The process for the treatment of zirconium based metal waste of claim 26 further comprising the step of storing said sintered green body. 29. A process for the treatment of zirconium based metal waste, the process including the steps of converting at least some of said ziconium based metal to an oxide, thereafter forming a mixture of said oxide with a sol which gel and is freeze-castable, producing a green body from the mixture, and sintering said green body. 30. A process according to claim 29 wherein the sol comprises a silica sol or a zirconia sol, and the mixture thereof with the said oxide is solidified by freeze-casting to form a solid, handleable green body for sintering. 31. A process according to claim 29 wherein filler powder is included in said mixture. 32. A process according to claim 29 wherein said mixture is frozen at a temperature in the region of -50.degree. C. 33. A process according to claim 29 wherein the green body is thawed prior to sintering. 34. The process for the treatment of zirconium based metal waste of claim 29 further comprising the step of storing said sintered green body.
053717694	description	PREFERRED EMBODIMENT OF THE INVENTION An embodiment of the present invention will be explained with reference to FIGS. 1 and 2. In these figures, the reference numeral 1 refers to straps of a thin strip form made of zircalloy. A large number of grid cells 2 of the grids are formed by cross assembling the straps 1 with each other. A plurality of bosses 3, arranged in rows and projecting out of one wall of the grid cells 2, are formed on both ends (the top and bottom ends in FIG. 1) of the straps 1. In the spaces between the two columns of bosses 3 are disposed rectangular-shaped central windows 4. As shown in FIG. 2, punch-out windows 5 (window sections) of a cross shape, consisting of an interlock section 5a and a recess section 5b at right angles to each other, are formed above and below and between the central windows 4. Near the center region of the strap 1, a series of through holes 6 of roughly rectangular shape are formed on the extension line joining the vertical pairs of window sections 5. At the top and bottom edges of the straps 1 on the extension line joining the vertical pairs of window sections 5 are disposed a series of vertical pairs of strap welding tabs 7. With regard to the first type of strap 1 as shown in FIG. 2, on one edge (i.e., the bottom strap welding tabs 7 and window sections 5 in FIG. 2) of the strap 1 is formed a series of slits 8 which start at the bottom strap welding tab 7 and pass through the lower window section 5 to reach the center region of the strap 1. The second type of the strap 1 has the same configuration as the first type except that a series of slits are formed from the top strap welding tab 7. When the first and second types of straps 1 are assembled at right angles by inserting each other at the slits 8, the intersections 9 are formed as shown in FIG. 1. A ladder-like spring frame member 10 made of Inconel is engaged with strap 1 by engaging at window sections 5 formed on the straps 1. The spring frame member 10 is disposed on the strap 1 at a height of central windows 4, and comprise: spring sections 11 jutting out oppositely to the bulging direction of the bosses 3; and a pair of connecting plate sections 12 which connect to the top and bottom sections of the spring section 11. Protrusion sections 13 which protrude in the opposite direction to the spring section 11 are formed on the connecting plate section 12 to be engaged with the interlock section 5a of the window section 5. Further, vertically opposing pairs of frame member welding tabs 14 are formed at the top and the bottom edges of the spring frame member 10. With regard to the first type of the spring frame member shown in FIG. 2, lower horizontal frame member welding tabs 14 are provided with slits 15. The second type of the spring frame member (not shown) is provided with slits at upper horizontal frame member welding tabs 14. When the first and second spring frame members are assembled, they are inserted to each other through the slits 15. The frame member welding tabs 14 are assembled with each other at right angles by inserting through the slits 15, a welding section 16 is formed at the intersection of the frame member welding tabs 14, as illustrated in FIG. 1. The welding section 16 is housed in the over-sized recess section of the window section 5 so as to prevent adverse thermal effects on the strap 1 during the welding operation of spring frame members 10. The method of constructing a grid using the strap configuration presented above will be explained with reference to FIG. 2. First, the spring frame members 10 are attached to the strap 1 by engaging the protrusion sections 13 of the spring frame members 10 in each interlock section 5a of the two rows of window sections 5. The straps 1 with the spring frame members 10 attached thereto are then cross-inserted through the slits 8 to construct a temporary structure shown in FIG. 1. Next, the strap welding tabs 7 crossing at the intersections 9 of the straps 1 are laser welded (exemplified by reference numeral 20 in FIG. 1), and the frame member welding tabs 14 of the spring frame member crossing at the welding section 16 are similarly laser welded. In this way, the grid assembly structure itself is constructed with the mutually crossing straps 1 made of zircalloy while only the spring sections 11 formed with mutually crossing spring frame members 10 made of Inconel are disposed inside the grind 2. Before loading fuel rods into the grids of the above construction, keys (not illustrated) are inserted thorough the through hole 6 horizontally into the grid cells 2 to press open the spring section 11, then fuel rods can be slid in the grid cells 2 without interference from the springs. Thenafter, the keys are removed to return the spring section 11 to the original position and retain the fuel rods with the spring sections 11 and the bosses 3.
	claims	1. A lighting system, in particular for EUV lithography, comprising a projection objective for producing semiconductor elements for wavelengths ≦193 nm, a light source, an object plane, an exit pupil, a first optical element having first grid elements for producing optical channels and a second optical element having second grid elements, each optical channel which is formed by one of the first grid elements of the first optical element being assigned a grid element of the second optical element, it being possible for grid elements of the first optical element and of the second optical element to be configured in such a way or arranged in such a way that the result for each optical channel is a continuous beam course from the light source as far as the object plane, characterized in that the angles of the first grid elements of the first optical element can be adjusted in order to modify a tilt in order, by means of tilting the first grid elements, to implement a different assignment of the first grid elements of the first optical element to the second grid elements of the second optical element. 2. The lighting system as claimed in claim 1, characterized in that the number M of second grid elements of the second optical element is greater than the number N of first grid elements of the first optical element. 3. The lighting system as claimed in claim 1 or 2, characterized in that the location and/or the angle of the second grid elements of the second optical element can be adjusted individually and independently of one another in order, by means of displacement and/or tilting of the first and second grid elements, to implement a different assignment of the first grid elements of the first optical element to the second grid elements of the second optical element. 4. The lighting system as claimed in claim 3, characterized in that the first grid elements are formed as field honeycombs in the form of first mirror facets, and in that the second grid elements are formed as pupil honeycombs in the form of second mirror facets, the first mirror facets and the second mirror facets in each case being arranged on a mirror support. 5. The lighting system as claimed in claim 4, characterized in that the optical channels between the mirror facets of the first and the second optical element can be adjusted by tilting the first mirror facets of the first optical element in relation to the mirror support, in order in this way to implement different assignments of the first mirror facets of the first optical element to the second mirror facets of the second optical element and therefore different illumination patterns of an exit pupil. 6. The lighting system as claimed in claim 4 or 5, characterized in that the optical channels between the first mirror facets of the first optical element and the second mirror facets of the second optical element can be adjusted by tilting and displacing the second mirror facets of the second optical element in relation to the mirror support. 7. The lighting system as claimed in claim 4, characterized in that the mirror facets of the first optical element and/or of the second optical element are in each case connected to the associated mirror support via a joint. 8. The lighting system as claimed in claim 7, characterized in that the joints are formed as solid body joints. 9. The lighting system as claimed in claim 7 or 8, characterized in that the mirror facets can be tilted in the x direction and/or in the y direction. 10. The lighting system as claimed in claim 9, characterized in that the joints are in each case located on the x axis and/or the y axis of the mirror facets. 11. The lighting system as claimed in claim 4, characterized in that, in order to displace and/or tilt the mirror facets, actuators are arranged between the grid elements and the mirror support. 12. The lighting system as claimed in claim 11, characterized in that the actuators have piezoceramic adjusting elements. 13. The lighting system as claimed in claim 12, characterized in that the actuators are provided with actuating elements that can be activated magnetically or electrically. 14. The lighting system as claimed in claim 11, characterized in that the actuators adjust the grid elements continuously via a control loop. 15. The lighting system as claimed in claim 11, characterized in that end positions are defined for the actuators. 16. The lighting system as claimed in claim 4, characterized in that the mirror facets can be displaced on predefined paths. 17. The lighting system as claimed in claim 16, characterized in that cam tracks, in which the mirror facets are guided individually in each case, are introduced into the mirror support. 18. The lighting system as claimed in claim 17, characterized in that the mirror support is formed as a guide disk, which interacts with a control disk, in which there are arranged guide tracks for the displacement of the mirror facets. 19. The lighting system as claimed in claimed 18, characterized in that the control disk is driven. 20. The lighting system as claimed in claimed 17, characterized in that each mirror facet is guided in a cam track in the mirror support, and in that each mirror facet can be driven individually by a drive element. 21. The lighting system as claimed in claim 20, characterized in that the drive element is in each case arranged in a cam track and each mirror facet is moved individually in accordance with the inch-worm principle. 22. A projection exposure installation for microlithography for producing semiconductor elements, comprising a lighting system and comprising a projection objective for producing semiconductor elements for wavelengths ≦193 nm, a light source, an object plane, an exit pupil, a first optical element having first grid elements for producing optical channels and a second optical element having second grid elements, each optical channel which is formed by one of the first grid elements of the first optical element being assigned a grid element of the second optical element, it being possible for grid elements of the first optical element and of the second optical element to be configured in such a way or arranged in such a way that the result for each optical channel is a continuous beam course from the light source as far as the object plane, characterized in that the angles of the first grid elements of the first optical element can be adjusted in order to modify a tilt in order, by means of tilting the first grid elements, to implement a different assignment of the first grid elements of the first optical element to the second grid elements of the second optical element. 23. The projection exposure installation as claimed in claim 22, characterized in that the number M of second grid elements of the second optical element is greater than the number N of first grid elements of the first optical element. 24. The projection exposure installation as claimed in claim 22 or 23, characterized in that the location and/or the angle of the second grid elements of the second optical element can be adjusted individually and independently of one another in order, by means of displacement and/or tilting of the first and second grid elements, to implement a different assignment of the first grid elements of the first optical element to the second grid elements of the second optical element. 25. The projection exposure installation as claimed in claim 24, characterized in that the first grid elements are formed as field honeycombs in the form of first mirror facets, and in that the second grid elements are formed as pupil honeycombs in the form of second mirror facets, the first mirror facets and the second mirror facets in each case being arranged on a mirror support. 26. The projection exposure installation as claimed in claim 25, characterized in that the optical channels between the mirror facets of the first and the second optical element can be adjusted by tilting the mirror facets of the first optical element in relation to the mirror support, in order in this way to implement different assignments of the first mirror facets of the first optical element to the second mirror facets of the second optical element and therefore different illumination patterns of an exit pupil. 27. The projection exposure installation as claimed in claim 25 or 26, characterized in that the optical channels between the first mirror facets of the first optical element and the second mirror facets of the second optical element can be adjusted by tilting and displacing the second mirror facets of the second optical element in relation to the mirror support.
051695944	summary	BACKGROUND OF THE INVENTION The present invention relates to a method of remotely installing or removing a nozzle dam of a nuclear stem generator system of the type disclosed in U.S. Pat. No. 5,032,350, assigned to the assignee of the present application. The disclosure of that patent, as well as the disclosures of U.S. Pat. Nos. 4,482,076; 4,744,392; 4,770,235 and 4,483,457 mentioned therein are hereby incorporated by reference. In performing the sliding assembly or disassembly operation of U.S. Pat. No. 5,032,350 it has been found useful to torque the segments about the mating surfaces which slide relative to each other to reduce the sliding friction force required. This is done to limit the segment edge face to edge face sliding contact and provide a slight gap therebetween. Thus, in assembly and disassembling it would be convenient to have a method to accomplish the sliding and described torquing operations from outside the head away from major radiation exposure. Accordingly, it has been desirable to have a method of facilitating all of the assembly and disassembly operations necessary to remotely install or remove a nozzle dam from its useful position, other than those performed by a manipulator as in U.S. Pat. No. 5,032,350 or its parent application, U.S. Pat. No. 4,954,312. The ideal method would include a procedure to assemble and disassemble the dam subassembly with slidlingly assembled segments; to secure or release rib mounted spring loaded camlock components for securement or detachment of the segments; to install and tension a tabbed seal diaphragm on the dam subassemble or remove same; and, to connect and disconnect push-pull operating quick-connects to connect or disconnect the subassembly mounted remotely actuated radial dam locking pins, fluid lines; all without entry of the head of the steam generator with attendant radiation exposure. SUMMARY OF THE INVENTION The invention is a method of remotely installing or removing a nozzle dam in a nuclear steam generator. The steam generator has at its lower end, a head, a manway penetrating the head, a nozzle penetrating the head and a manipulator having a free end in the head. The manipulator includes a clamp mechanism for holding and orienting an inner or center nozzle dam segment while outer nozzle dam segments are secured thereto or detached, at least partially, therefrom by a sliding or a hinged action. The dam segments are sized to pass through the manway and each has means for engaging its adjacent segments thereon for assembly and secured by components into a dam subassembly which can also be disassembled by the method. A Titan 7F manipulator or the like is mounted and arranged for orienting and translating the dam subassembly within the head to move the dam subassembly to its seat within the nozzle where it is to be locked and radially secured therein by subassembly mounted remotely activated radial pins. The manipulator also is used to move the dam from the seat after it is unlocked for removal from the seat and disassembly. The novel method of the invention is for use in the just described broad method and is the specific procedure which involves providing and using a tool set having a plurality of elongated tools for engagement and disengagement with the dam segments and components for securement during assembly and for disassembly. This includes manually operating the tools in association with the manipulator from outside the head for operator protection and ease of use of the manipulator. The method includes the step of physical manipulation of the elongated tools of the set from outside the head to minimize exposure of the tool operator to radiation from within the head and to avoid the necessity of the operator to enter the head while: 1. aligning the segments relative to each other for movement into or out of assembled position during manipulation of the elongated tools; and, PA1 2. securing or detaching the segments in or from proper aligned position to form or disassemble a dam subassembly secured by components of the dam subassembly. The aligning of the segments is a method which includes using an elongated bifurcated tool of the set to engage a segment edge and torquing one segment relative to the other to relieve edgewise sliding frictional surface contact by a resulting alignment of the segments. The novel procedure for securing or detaching the segments is a method which includes using an elongated tool with a double offset T-shaped end to operate segment rib mounted spring loaded camlock components of the type disclosed in U.S. Pat. No. 4,744,392 "Nozzle Dam Segment Bolt and Keeper". The method includes use of yet another elongated tool which can be manipulated from outside the head over and between the ribs of the segments to operate quick-connects in fluid lines to subassembly mounted remotely actuated radial pins including fluid lines, for example, to disconnect the fluid lines during disassembly of the segments for removal from the head. This tool includes an offset and transverse body-straddling bifurcated end for engaging the end of a quick-connect sleeve. The offset provides a portion with an axis substantially parallel to the stem and body, such that by engagement and physical manipulation of the bifurcated end against the quick connect sleeve, from the handle end at the opposite end of a shaft portion, the offset bifurcation pulls back the body sleeve to accomplish disconnection. The typical quick-connect used is a SWAGELOK.RTM. "Keyed" quick-connect, Model QC6 (KS Blue), with the ball check removed for free flow. It is available from The Crawford Fitting co. of Solon, Oh. Another elongated tool used in the method of the invention facilitates tensioning of the diaphragm seal tabs to install or remove the diaphragm from the dam subassembly. The tabs extend from the diaphragm periphery and preferably have an opening for hooked engagement with projections, typically screws, spaced on the subassembly peripheral margin to face the inside of the head. The tabs are tensioned for hooking and unhooking operations by providing and using an elongated tool from the tool set. It is characterized by an elongated shaft portion with a transversely extending hook end and an opposite handle end for manipulation from outside the head. The tensioning includes the step of inserting the hook end in a tab opening to provide hooking or unhooking clearance within that tab opening or an adjacent tab opening for movement of the tab into or out of hooked engagement with one of the projections. Preferably there are multiple tab openings in the tab and the opening engaged by the hooked end of the tool to which the tension is applied, is closer to the tab end than the tab opening that creates the hooked tension connection of the diaphragm to the subassembly. The invention, then is the method of using the tools to remotely install and/or remove a nozzle dam without entering the head.
048658014	summary	FIELD OF THE INVENTION The present invention relates to a shielding device for use in shielding the under-vessel area from radiation emanating above the lower terminus of a control rod drive of a power generating nuclear reactor of the boiling water type. BACKGROUND OF THE INVENTION The nuclear reaction in a power generating nuclear reactor of the boiling water type, also referred to as a boiling water reactor, is controlled by a set of hydraulically actuated bottom-entry control rods. The control rod drive pistons enter the pressure vessel of the reactor through vertical penetrations in the lower plenum of the reactor known as control rod drive housings. These housings extend below the pressure vessel of the reactor far enough to provide the pistons sufficient travel so that the control rods may be completely withdrawn from the reactor fuel core. At the bottom of each housing, the hydraulic actuation mechanisms for each piston are bolted to the housings with flange connections. These flange connections are typically secured by eight socket head capscrew bolts that are symmetrically located around the control rod drive flange. The hydraulic actuation lines, sometimes referred to as insert/withdraw lines, enter the flange assembly through ports which are typically located on the top surface of the control rod drive housing flange. Control room reactor operators must know the vertical position of each control rod at all times. Magnets located on the bottom of the control rod drive pistons, when sensed by an electronic probe, provide this information to the operator for each control rod. These electronic sensors, or position indicating probes, send their signals to the control room through the probe cables which are plugged into probe housings at the bottom of each control rod drive. The flange assemblies contain a number of hydraulic components which relate to the actuation of the control rod drive pistons. During use, reactor coolant is employed as the working fluid flowing through these hydraulic components. Particulate debris consisting of neutron-activated corrosion products and wear fragments, as well as the water soluble radioactive contaminants that result from normal plant operations tend to accumulate in the hydraulic system. The particulate debris is from the alloys that are present in the various piping systems and internal components in a boiling water reactor. The water soluble radioactive contaminants carried by the reactor coolant through the flange assemblies are either fission products which have leached through the small perforations of failed fuel rods or materials that become radioactive through neutron-activation. None of the materials used in the piping or other reactor structural components are radioactive when they are installed. However, as a result of normal use, these structures and components will tend to accumulate radioactive contaminants, and will eventually become sources of radiation themselves. Many of the materials used in the piping and structural components are subject to corrosion, friction and wear, and these processes release metallic particulates into the reactor coolant. The reactor coolant carries these particulates throughout the entire system. While the reactor is operating, a very dense flux of neutrons is present in the reactor's fuel core, and the particulates and contaminants carried by the coolant are exposed to this neutron flux. When a high-energy neutron collides with an atom of any material, the atom is typically transmuted into a new species, and in nearly all cases the transmuted atom is unstable or radioactive. This process is known as neutron-activation. It is through neutron activation such as this that the particulates and contaminants in the reactor coolant are made radioactive. Eventually, the particulates and contaminants in the coolant accumulate in regions of slow flow or plate-out onto attractive surfaces which are contaminated with radioactivity in this manner. Because these particulates and contaminants are often carried through the reactor fuel core many times before settling or plating-out, nearly all of them have experienced neutron activation. It is through this cycle of (1) corrosion, friction or wear related deposition into the reactor coolant; (2) neutron activation; and (3) sedimentation, accumulation or plate-out that the piping and other reactor components eventually become contaminated with radioactive materials. The neutron activation process in a nuclear reactor of the boiling water type produces a great number of radioactive nuclides. Most of these radioactive nuclides are derived from the constituent elements of stainless-steel or any other alloy present, such as brass, if brass is present in the condenser tubing of the reactor. Typical nuclides are the several radioactive isotopes of Cobalt, Copper, Iron, Manganese, Nickel, Tin and Zinc. The radioactive water soluble contaminants will typically consist mostly of the various radioactive isotopes of Cesium and Iodine. However, in practice, one single nuclide, Cobalt-60 or Co-60, dominates the total radiation produced by all others to such a degree that practical experience has shown that shielding can be designed substantially as if that were the only nuclide present. Cobalt is often used in industrial applications to increase the toughness of various alloys. Essentially all of the naturally occurring Cobalt is the stable and non-radioactive isotope Co-59. When Co-59 is carried by the Reactor Coolant through the fuel core the neutron activation process typically yields the radioactive nuclide Co-60. Decay of Co-60 emits radiation in the form of a simultaneous pair of gamma photons with energies of 1.17 MeV and 1.33 MeV, and its half-life is roughly 5.25 years. Metallic lead is an excellent shielding material for this type of radiation. As these radioactive materials accumulate on the interior surfaces of the reactor vessel, the control rod drive housing and the flange assemblies, these components themselves become intense sources of radiation of all forms, i.e., alpha particles, beta particles and gamma photons. In addition, the outer surfaces of these components can become contaminated from other sources, such as the release of radioactive contaminants caused by the deluge of water from an adjacent control rod drive. In practice, however, only gamma radiation, dominated by the two high-energy Co-60 photons, can penetrate the thick stainless-steel walls of the flange assembly. The resulting contact dose rates for these components, even months after reactor shutdown, routinely range as high as ten Roentgen Equivalent Man (10 rem) per hour and can be substantially higher in certain cases. A Roentgen Equivalent Man (rem) is the most common unit employed to measure radiation dose rates and is usually defined as that quantity of any type of ionizing radiation which when absorbed by man, produces an effect equivalent to the absorption by man of one Roentgen of X- or gamma radiation. A Roentgen is that quantity of X- or gamma radiation such that the associated corpuscular emission per 0.001293 grams of air produces, in air, ions carrying one electrostatic unit of quantity of electricity of either sign. Because of the close proximity of the flange assemblies to the under-vessel workers doing routine maintenance and replacement work, radiation dose rates may range from 0.3 to 0.7 rem per hour to the whole body, which is defined as the trunk, head and lens of eyes, of each worker. The U.S. Nuclear Regulatory Commission has issued regulations governing the operation of all nuclear facilities. These regulations appear at Title 10 of the Code of Federal Regulations. In Part 20, Section 101, Paragraph (a), 10 CFR 20.101(a), the regulations currently restrict occupational doses to the whole body of a worker to a standard of 1.25 rem per calendar quarter, although the concept of `dose banking` as that term is used in 10 CFR 20.101(b), presently permits worker exposure of up to 3.00 rem per calendar quarter under very restrictive circumstances. However, the philosophy of "As Low As Reasonably Achievable", or ALARA, as it is embraced by both the Nuclear Regulatory Commission and the nuclear industry in general, motivates most reactor operators to limit occupational doses to much lower levels. Routine under-vessel maintenance performed during boiling water reactor refueling outages can typically require from 600 to over 1000 man-hours to complete. Unfortunately, the large dose rates which can be caused by the close proximity of the flange assemblies make it possible for an under-vessel worker to receive his entire quarterly dose allowance in just a few hours. Such workers, who have used their allowable dose limit, are not permitted to work in areas where they may be exposed to any further radiation. Under-vessel maintenance work, as currently practiced, is not only a radiological safety concern, it is also a very serious manpower management problem as well. OBJECT OF THE INVENTION It is an object of the invention to provide a shield device for shielding the lower terminus of a control rod drive in a power generating nuclear reactor of the boiling water type. It is another object of the invention to provide a method for reducing the exposure of workers to radiation during under-vessel maintenance and replacement work. The other objects, features and advantages of the present invention will become more apparent in light of the following detailed description of the preferred embodiment thereof. According to the present invention, there is provided a shielding device for us in shielding the under-vessel area from radiation emanating from above the lower terminus of a control rod drive of a power generating nuclear reactor of the boiling water type, which control rod drive has a flange assembly comprised of a control rod drive housing flange and a mating control rod drive flange joined together by bolting means, which shielding device comprises: an effective amount of shielding material having a suitable geometry, and means to hold said shielding material in position shielding the under-vessel area from said radiation. According to another embodiment of the present invention, there is provided a method for reducing the exposure of workers to radiation emanating from above the lower terminus of a control rod drive during under-vessel maintenance and repair work on a power generating nuclear reactor of the boiling water type, which method comprises: (a) attaching to the lower terminus of each control rod drive, a shielding device which comprises an effective amount of shielding material having a suitable geometry, and means to hold said shielding material in position shielding the under-vessel area from said radiation; (b) performing the maintenance or repair work; and (c) removing said shielding device from each control rod drive. According to a preferred embodiment of the present invention, there is provided a shielding device, for use in shielding the under-vessel area from radiation emanating from above the lower terminus of a control rod drive of a power generating nuclear reactor of the boiling water type, which control rod drive has a flange assembly comprised of a control rod drive housing flange and a mating control rod drive flange joined together by bolting means, which shielding device comprises: a plurality of shielding members wherein (a) each said shielding member is comprised of: (i) a cylindrical outer wall of a non-corrosive metallic substance, said outer wall having a vertical length which extends from a point just above the control rod drive flange assembly downward for a suitable distance, and which outer wall has an inside radius of curvature substantially equal to the outside radius of said control rod drive flange assembly, (ii) an annular outer bottom wall of a non-corrosive metallic substance, said outer bottom wall joining and fixedly attached to said outer wall at the lower periphery of said outer wall, said outer bottom wall having an opening therethrough, which is central to the complete annulus, (iii) a first cylindrical inner wall of a non-corrosive metallic substance, said first inner wall extending downward from a point just below the flange assembly, said first inner wall member having an outside radius of curvature smaller than the radius of curvature of the outer cylindrical wall by an amount great enough to define a partial cylindrical chamber therebetween, which partial cylindrical chamber is large enough to accommodate an effective amount of shielding material therein, (iv) an annular inner bottom wall of a non-corrosive metallic substance fixedly attached to said first inner wall, said inner bottom wall disposed above said outer bottom wall by a distance great enough to define a partial annular chamber, which partial annular chamber is large enough to accommodate an effective amount of shielding material therein, said inner bottom wall having an opening therethrough, which opening is central to the complete annulus, (v) an annular upper interior wall of a non-corrosive metallic substance fixedly attached to the interior surface of said outer wall and the upper periphery of said first inner wall, thereby defining a top to said partial cylindrical chamber, (vi) a second cylindrical inner wall of a non-corrosive metallic substance fixedly attached to the interior periphery of said outer bottom wall and said inner bottom wall, thereby defining an interior wall of said partial annular chamber, said second cylindrical inner wall partially defines a central opening of the complete annulus which opening is of large enough dimension to accommodate any control cables or connectors, (vii) two radial walls of a non-corrosive metallic substance, which radial walls enclose the radial edges of said partial cylindrical chamber and said partial annular chamber, and are fixedly attached to the second cylindrical inner wall, the outer bottom wall, the inner bottom wall, the first inner cylindrical wall, the upper interior wall and the outer wall, thereby enclosing both the partial cylindrical chamber and the partial annular chamber, (viii) an effective amount of shielding material disposed within said partial cylindrical chamber, and (ix) an effective amount of shielding material disposed within said partial annular chamber; (b) each said shielding member is provided with attachment means comprising: (i) a first contact point which comprises an inward-facing partial rim or lip fixedly attached to the upper periphery of said outer wall; and (ii) a second contact point which comprises an upwardly adjustable wedge-shaped member which can be positioned on the under side of the control rod drive flange, interior to said bolting means; and (c) each said shielding member has a suitable geometry such that, when said plurality of shielding members are assembled in mating contiguous engagement, they surround and enclose the lower terminus of said control rod drive with an effective amount of shielding material.
	summary
	description	This application is filed under the provisions of 35 U.S.C. §371 and claims the priority of International Patent Application No. PCT/US2011/046282 filed on 2 Aug. 2011 entitled “ELECTRON MICROSCOPE SAMPLE HOLDER FOR FORMING A GAS OR LIQUID CELL WITH TWO SEMICONDUCTOR DEVICES” in the name of John Damiano, Jr., et al., which claims priority of U.S. Provisional Patent Application No. 61/369,772 filed on 2 Aug. 2010, all of which are hereby incorporated by reference herein in their entirety. The invention relates generally to sample holders used for mounting samples in an electron microscope, e.g., a transmission electron microscope (TEM), a scanning transmission electron microscopy (STEM) and variations of the scanning electron microscopes (SEM) that use traditional TEM-type holders and stages, for imaging and analysis. The sample holder is a component of an electron microscope providing the physical support for specimens under observation. Sample holders traditionally used for TEMs and STEMs, as well as some modern SEMs, consist of a rod that is comprised of three key regions: the end (300), the barrel (200) and the sample tip (100) (see, e.g., FIG. 1). In addition to supporting the specimen, the sample holder provides an interface between the inside of the instrument (i.e., a vacuum environment) and the outside world. To use the sample holder, one or more samples are first placed on a support device. The support device is then mechanically fixed in place at the sample tip, and the sample holder is inserted into the electron microscope through a load-lock. During insertion, the sample holder is pushed into the electron microscope until it stops, which results in the sample tip of the sample holder being located in the column of the microscope. At this point, the barrel of the sample holder bridges the space between the inside of the microscope and the outside of the load lock, and the end of the sample holder is outside the microscope. To maintain an ultra-high vacuum environment inside the electron microscope, flexible o-rings are typically found along the barrel of the sample holder, and these o-rings seal against the microscope when the sample holder is inserted. The exact shape and size of the sample holder varies with the type and manufacturer of the electron microscope, but each holder contains these three key regions. The sample holder can also be used to provide stimulus to the sample, and this stimulus can include temperature, electrical current, electrical voltage, mechanical strain, etc. One type of sample is a semiconductor device. The semiconductor device can be designed to have an array of electrical contact pads on it, and the sample holder can be designed to transfer electrical signals from an external source, through the holder, to the semiconductor device. The need for high density arrays comes from an increasing demand to perform a wider variety of experiments on a sample within the microscope—a field known as in situ microscopy. As demonstrated in previous filings by the present inventors, semiconductor devices can be made to interact with a specimen positioned thereon by passing current or creating fields across or near a specimen. These electrical signals can be used to heat, cool, bias or charge a specimen, all while being viewed in real time within the microscope. Increasing the number of electrical contacts provided to a specimen increases the number of or type of experiments that can be done on the same device. One type of sample holder is one in which (1) two semiconductor devices can be placed, (2) specimens can be placed on or near the semiconductor devices and (3) using the combination of holder and devices, the specimen's environment, including an electrical field and a gas or liquid flow, can be precisely controlled. Further, methods to introduce liquid or liquid mixtures to specimens within the microscope while containing and controlling the environment around specimens have been developed. However, there is a need for more advanced apparatuses and methods to contact and align devices used to form liquid or gas cells. The present invention relates generally to a novel liquid or gas sample holder which provides improved contact and alignment for microelectronic devices in the sample holder and as well as electrical contacts and multiple liquid or gas inputs/outputs to the specimens or microelectronic devices. In one aspect, a sample holder for an electron microscope is described, said sample holder comprising a sample holder body and a sample holder lid, wherein the sample holder body comprises at least one pocket having a pocket bottom and pocket walls for the positioning of two microelectronic devices therein, and the sample lid has a top side and a bottom side. In another aspect, a method of imaging a sample in a liquid and/or gaseous environment in an electron microscope is described, said method comprising inserting a sample in a sample holder, inserting the sample holder comprising the sample in an electron microscope, introducing a liquid and/or gas to the sample in the sample holder, and imaging the sample in the electron microscope, wherein the sample holder comprises a sample holder body and a sample holder lid, wherein the sample holder body comprises at least one pocket having a pocket bottom and pocket walls for the positioning of two microelectronic devices therein, and the sample lid has a top side and a bottom side. Other aspects, features and embodiments of the invention will be more fully apparent from the ensuing disclosure and appended claims. The present invention generally relates to novel samples holders, methods for introducing liquids or gases to the sample holder, and uses of the novel sample holder. It is to be understood that the specimen holder and specimen holder interface described herein are compatible with and may be interfaced with the semiconductor specimen support devices disclosed in International Patent Application Nos. PCT/US08/63200 filed on May 9, 2008, which is incorporated herein by reference in its entirety. It should be appreciated by one skilled in the art that alternative semiconductor sample support devices may be interfaced with the sample holder described herein. The sample holder provides mechanical support and a liquid or gaseous environment for one or more specimens and/or semiconductor support devices and may also provide electrical contacts to the specimens and/or semiconductor support devices. The sample holder can be manufactured with tips, barrels and ends of various shapes and sizes such that the specimen holder fits any manufacturer's electron microscope. As defined herein, a “membrane region” on the semiconductor device corresponds to unsupported material comprised, consisting of, or consisting essentially of carbon, silicon nitride, SiC or other thin films generally 1 micron or less having a low tensile stress (<500 MPa), and providing an at least partially electron transparent region for supporting the at least one specimen. The membrane region may include holes or be hole-free. The membrane region may be comprised of a single material or a layer of more than one material and may be either uniformly flat or contain regions with varying thicknesses. As defined herein, “microelectronic” means a semiconductor material, such as silicon, that is intermediate in electrical conductivity between conductors and insulators. As defined herein, “device” means a structure used to either contain liquids or gases around a specimen and includes, but is not limited to, a window device, an electrical device and a heating device. As defined herein, a “cell” corresponds to a region defined by two substantially parallel positioned devices, wherein at least one liquid and/or gas can be flowed therethrough. A specimen can be positioned within the cell for imaging purposes. As defined herein, “specimen” means the object being studied in the electron microscope, typically placed within or on the device in the region of liquid or gas control which is at least partially electron transparent (e.g., nanoparticle, catalyst, thin section, etc.). As defined herein, a “pocket” corresponds to a space in the sample cell holder that defines the vertical walls of the cell, into which the two substantially parallel devices are positioned to form the cell. As defined herein, “contact points” correspond to protrusions from the walls of the pocket that are engineered to align the devices when positioned in the pocket. As defined herein, “window device” means a device used to create a physical, electron transparent barrier on one boundary and the vacuum environment of the electron microscope on the other and is generally a silicon nitride-based semiconductor micro-machined part, although other semiconductor materials are contemplated. As defined herein, “frame” means a rigid region around the perimeter of a device that is used to provide mechanical support to the entire device structure. Preferred embodiments include a silicon frame, even more preferably a silicon frame selectively etched using KOH, a silicon frame selectively etched using reactive ion etching (RIE), a silicon frame selectively etched using deep reactive ion etching (DRIE), or a silicon frame released from an silicon-on-insulator (SOI) wafer. The present application improves on the prior art by providing a simple apparatus and method for aligning and exchanging devices and making electrical contacts to said devices. A schematic of a generic window device is shown in FIG. 2. A thin membrane region, e.g., amorphous silicon nitride, forms the window whereby imaging and analysis can be performed through the window. The window must withstand a differential pressure between the gas cell and the vacuum environment of the microscope. The window's “frame” is preferably single-crystal silicon. The frame is formed by selectively etching a cavity in the single-crystal silicon substrate. A thin “spacer” layer can be formed around the membrane window (for example as shown in the window device in FIG. 6). The thickness of this layer can be precisely set, and, when a second device, e.g., a heating device or another window device, is stacked atop the window device, the thickness of the spacer sets the distance between the devices and hence the thickness of the gas or liquid layer between the devices. Preferred spacer thickness is in a range from about 0.1 μm to about 50 μm. Spacer materials contemplated herein include, but are not limited to, epoxy-based photoresists such as SU-8 (Microchem, Newton, Mass.), grown or deposited semiconductor layers, deposited or electroplated metal films and polyimide films such as the HD-4100 series of polymers (Hitachi Dupont MicroSystems LLC). A schematic of a generic electrical biasing device is shown in FIG. 3. The electrical biasing device has electrodes that run from the edge of the device to the center of a thin silicon nitride membrane. Specimens can be placed on the silicon nitride membrane region for inspection. Typically voltage or current is applied to the electrodes at the edge of the chip, and these signals travel to the membrane region and the specimen. The “frame” portion of the device, surrounding the membrane, can be single-crystal silicon. The frame is formed by selectively etching a cavity in the single-crystal silicon substrate. Gold contact pads are used to form the electrodes. The silicon nitride material is electrically insulating. A thin “spacer” layer can be formed around the membrane window. The thickness of this layer can be precisely set, and, when a second device, e.g., a window device, is stacked atop the electrical device, the thickness of the spacer sets the distance between the devices and hence the thickness of the liquid layer between the devices. Preferred spacer thickness is in a range from about 0.1 μm to about 50 μm. For example, the spacer layer can be removed over the gold electrodes at the edge of the electrical device where contacts are formed. The cut in the spacer layer forms a seal around the contact when the devices are stacked and prevents the liquid from reaching the contact point between the device and the sample holder. It should be appreciated that the electrical biasing device can be larger, smaller, or the same dimensions as the window device. A schematic of a generic heating device is shown in FIG. 4. Specimens can be placed on the thin membrane region, which is formed from layers of a conductive ceramic material, e.g., silicon carbide. When electrical current is forced through the ceramic membrane, the membrane region heats, heating the specimen. The “frame” portion of the device, surrounding the membrane, can be single-crystal silicon. The frame is formed by selectively etching a cavity in the single-crystal silicon substrate. Gold contact pads are used to form electrical contacts to the ceramic material. An electrically insulating layer of silicon dioxide or equivalent thereof between the ceramic layers and the silicon substrate prevents current flow from the ceramic membrane to the substrate, so all current stays in the membrane. In the embodiment shown in FIG. 4, the gold contact pads extend to one side of the device. It should be appreciated that the heating device can be larger, smaller, or the same dimensions as the window device. An electrochemical cell holder comprises a window device and an electrical device. A thermal cell holder comprises a window device and a heating device. The liquid cell holder uses two window devices. At least one device includes a spacer layer. When one device is placed atop the other, with the nitride membranes facing each other, the gap between the devices is set by the thickness of the spacer layer(s). Controlling the thickness of this spacer layer is critical: if the spacer is too thin, the specimens will get crushed. If the spacer is too thick, then there can be extra liquid around the specimen, which degrades the resolution of the image. The thickness of the spacer layer(s) is readily determined by the skilled artisan. FIG. 5 shows a cross-sectional schematic of the complete electrochemical cell with electrical biasing and window devices included. Liquid supply lines are not shown for clarity. The electrochemical cell is comprised of an electrochemical cell holder body, an electrochemical cell holder lid, one window device, one electrical biasing device, and o-rings used to seal the cell. The devices are stacked such that the membrane portion of each device is aligned. Additionally, the pair of stacked membranes is also aligned with holes in the holder body and holder lid such that an electron beam can pass through the entire assembly. FIGS. 7 and 8A-8C show cross-sectional and plan view schematics of the electrochemical cell holder and demonstrate how devices are loaded into the sample holder, how electrical contacts are formed, and how the environment inside the electrochemical cell is maintained. FIG. 7 shows the empty holder body and holder lid structures with a cross-sectional view of both. The holder body can have a cavity with a deep pocket and a shallow pocket when the size of the electrical or thermal device is different from that of the window device (e.g., in FIG. 7, the window device is smaller in length than the electrical or thermal device although it is contemplated herein that the electrical or thermal device may be smaller in length than the window device). It should be appreciated that when the window device and the electrical or thermal device have the same length and width that the holder body can have one deep cavity for accommodating both devices. The deep pocket has a bottom with a electron beam hole roughly centered in the pocket, and at least one o-ring or other sealing means can be placed around the hole. The depth of the pocket relative to the shallow pocket plane is approximately the thickness of the window device. The length and width of the deep pocket is slightly larger than a window device, as will be discussed at length hereinbelow. The length and width of the shallow pocket is slightly larger than the electrical or thermal device, as will be discussed hereinbelow. The shallow pocket fully encloses the deep pocket. On one side of the shallow pocket, away from the deep pocket, a row of contact points can be positioned that are used to electrically contact the electrical or thermal device when this device is loaded into the holder body. The depth of the shallow pocket is approximately the thickness of the electrical or thermal device. The holder lid can have a thick region and a thin region for alignment convenience. The length and width of the thick region is approximately the size of the shallow pocket in the holder body, such that the thick region can be inserted into the shallow pocket. An electron beam hole is placed in roughly the center of the thick region, and an o-ring or other sealing means surrounds the hole. The thin region extends beyond the thick region, and has more than one hole that allows the user to fasten the holder lid to the holder body with screws or other fastening means. A second o-ring or other sealing means surrounds the thick region and is used to form a seal between the holder lid and holder body. Gases or liquids are supplied to the pockets through supply lines in the holder body that extend from the pocket(s) to the outside of the opposite end of the holder. FIG. 8A shows the window device loaded into the holder body along the A-A′ cross-section for illustrative purposes. The device is placed in the deep pocket with the spacer layer (when present) facing up and the etched cavity in the substrate facing down. The bottom of the device sits on the o-ring or other sealing means surrounding the hole in the deep pocket, and a continuous seal is formed around the hole when pressure is applied and the o-ring is compressed between the holder body and the device. The top of the window device is at roughly the same height as the bottom plane of the shallow pocket. FIG. 8B shows the electrical biasing device loaded into the holder body for illustrative purposes. The device is placed in the shallow pocket with the spacer layer and/or thin membrane region facing down (i.e., facing the thin membrane region of the underlying window device) and the etched cavity in its substrate facing up. The spacer layer features on the electrical biasing device sit atop the window device. The gold contact pads on the electrical biasing device are aligned with the underlying contact points in the shallow pocket. Since the deep pocket and shallow pocket can have two common sides, as indicated in FIG. 7, the devices in the pockets are aligned to a common surface and are therefore aligned to each other. This critical “self-alignment” feature allows the devices to be aligned with respect to each other with great precision. FIG. 8C shows the final assembly of the electrochemical cell holder for illustrative purposes. The holder lid is placed atop the holder body and affixed to the holder body using more than one screw or other fastening means. The thicker portion of the lid fits atop the device stack. The o-ring or other sealing means in the thicker portion, which is positioned in proximity to the hole in the lid, fits around the cavity of the electrical device and forms a seal between the electrical device and the lid. The thickness of this thicker portion determines the amount of compression on the o-rings or other sealing means that surround the holes in the holder body and the holder lid. Adequate compression is required to form a tight seal around these holes and prevent gases or liquids from escaping around the devices. Attaching the holder lid pushes down on the device stack, compressing these o-rings or other sealing means and forming a seal. Adequate compression will also force the gold contact pads on the electrical device on to the underlying contact points, forming an electrical contact. The distance between the devices is set by the thickness of the spacer layer and this distance does not change when the devices are pushed together when sealing the cell. The second o-ring or sealing means on the holder lid, in the thinner portion of the lid, forms a seal between the holder body and the holder lid. This o-ring is compressed when the lid is attached to the body and prevents gases or liquids from escaping from between the holder body and the holder lid. FIG. 6 shows a cross-sectional schematic of the complete thermal cell with heating and window devices included. Gas or liquid supply lines are not shown for clarity. The thermal cell is comprised of a gas cell holder body, a gas cell holder lid, one window device, one heating device, and o-rings used to seal the cell. The devices are stacked such that the membrane portion of each device is aligned. Additionally, the pair of stacked membranes is also aligned with holes in the holder body and holder lid such that an electron beam can pass through the entire assembly. FIGS. 7 and 9A-9C show cross-sectional and plan view schematics of the thermal cell holder and demonstrate how devices are loaded into the sample holder, how electrical contacts are formed, and how the environment inside the thermal cell is maintained. FIG. 9A shows the window device loaded into the holder body along the C-C′ cross-section for illustrative purposes. The device is placed in the deep pocket with the spacer layer facing up and the etched cavity in the substrate facing down. The bottom of the device sits on the o-ring or other sealing means surrounding the hole in the deep pocket, and a continuous seal is formed around the hole when pressure is applied and the o-ring is compressed between the holder body and the device. The top of the window device is at roughly the same height as the bottom plane of the shallow pocket. FIG. 9B shows the thermal device loaded into the holder body for illustrative purposes. The device is placed in the shallow pocket with the spacer layer and/or thin membrane region facing down (i.e., facing the thin membrane region of the underlying window device) and the etched cavity in its substrate facing up. The thermal device sits atop the spacer layer features on the window device. The gold contact pads on the thermal device are aligned with the underlying contact points in the shallow pocket. Since the deep pocket and shallow pocket can have two common sides, as indicated in FIG. 7, the devices in the pockets are aligned to a common surface and are therefore aligned to each other. This critical “self-alignment” feature allows the devices to be aligned with respect to each other with great precision. FIG. 9C shows the final assembly of the thermal cell holder for illustrative purposes. The holder lid is placed atop the holder body and affixed to the holder body using more than one screw or other fastening means. The thicker portion of the lid fits atop the device stack. The o-ring or other sealing means in the thicker portion, which is positioned in proximity to the hole in the lid, fits around the cavity of the heating device and forms a seal between the heating device and the lid. The thickness of this thicker portion determines the amount of compression on the o-rings or other sealing means that surround the holes in the holder body and the holder lid. Adequate compression is required to form a tight seal around these holes and prevent gases or liquids from escaping around the devices. Attaching the holder lid pushes down on the device stack, compressing these o-rings or other sealing means and forming a seal. Adequate compression will also force the gold contact pads on the thermal device on to the underlying contact points, forming an electrical contact. The distance between the devices is set by the thickness of the spacer layer and this distance does not change when the devices are pushed together when sealing the cell. The second o-ring or sealing means on the holder lid, in the thinner portion of the lid, forms a seal between the holder body and the holder lid. This o-ring is compressed when the lid is attached to the body and prevents gases or liquids from escaping from between the holder body and the holder lid. Another embodiment of the gas or electrochemical cell is shown in FIGS. 15-18. In this embodiment, a schematic of an alternate generic window device with gold electrodes is shown in FIG. 15. Typically voltage or current is applied to the electrodes at the edge of the chip, and these signals travel to contact areas near the edge of the thin membrane region. The thin membrane region, e.g., amorphous silicon nitride, forms the window whereby imaging and analysis can be performed through the window. The window must withstand a differential pressure between the gas cell and the vacuum environment of the microscope. The window's “frame” is single-crystal silicon. The frame is formed by selectively etching a cavity in the single-crystal silicon substrate. Gold contact pads are used to form the electrodes. A thin insulating layer is formed around the membrane window. This layer prevents shorting between the device and the holder body when the device is placed into the holder body. Insulating materials contemplated herein include, but are not limited to, epoxy-based photoresists such as SU-8 (Microchem, Newton, Mass.), grown or deposited semiconductor layers and polyimide films such as the HD-4100 series of polymers (Hitachi Dupont Micro Systems LLC). A schematic of an alternate generic heating device is shown in FIG. 16. Specimens can be placed on the thin membrane region, which is formed from layers of a conductive ceramic material, e.g., silicon carbide. When electrical current is forced through the ceramic membrane, the membrane region heats, heating the specimen. The “frame” portion of the device, surrounding the membrane, can be single-crystal silicon. The frame is formed by selectively etching a cavity in the single-crystal silicon substrate. Gold contact pads are used to form electrical contacts to the ceramic material. An electrically insulating layer of silicon dioxide or equivalent thereof between the ceramic layers and the silicon substrate prevents current flow from the ceramic membrane to the substrate, so all current stays in the membrane. In the embodiment shown in FIG. 16, the thickness of the gold contact pads can be precisely set, and, when a second device, e.g., a window device, is stacked atop the heating device, the thickness of the heating device gold contact pads sets the distance between the devices and hence the thickness of the gas or liquid layer between the devices. Preferred spacer thickness is in a range from about 0.1 μm to about 50 μm. FIG. 17 shows a cross-sectional schematic of the complete alternate gas or electrochemical cell with heating and window devices included. Gas or liquid supply lines are not shown for clarity. The cell is comprised of a holder body, a holder lid, one window device, one heating device, and o-rings used to seal the cell. The devices are stacked such that the membrane portion of each device is aligned. Additionally, the pair of stacked membranes is also aligned with holes in the holder body and holder lid such that an electron beam can pass through the entire assembly. FIG. 18 shows cross-sectional and plan view schematics of the alternate gas or electrochemical cell holder body and holder lid without devices inserted. FIGS. 19A-19C show cross-sectional and plan view schematics of the gas or electrochemical cell holder and demonstrate how devices are loaded into the sample holder, how electrical contacts are formed, and how the environment inside the cell is maintained. FIG. 19A shows a thermal device loaded into the holder body for illustrative purposes to form a thermal cell. Alternatively, an electrical device could be loaded to form an electrochemical cell. The device is placed in the deep pocket with the gold contact pads facing up and the etched cavity in the substrate facing down. The bottom of the device sits on the o-ring or other sealing means surrounding the hole in the deep pocket, and a continuous seal is formed around the hole when pressure is applied and the o-ring is compressed between the holder body and the device. The top of the window device is at roughly at the same height as the bottom plane of the shallow pocket. A second o-ring is placed at the bottom of the shallow pocket surrounding the thermal device. FIG. 19B shows the window device loaded into the holder body for illustrative purposes. The device is placed in the shallow pocket with the thin membrane region facing down (i.e., facing the underlying window device) and the etched cavity in its substrate facing up. The window device sits atop the gold contact pad features on the thermal device and atop the o-ring that surrounds the thermal device. The gold contact pads on the thermal device are aligned with the contact areas on the window device, and the electrodes on the window device are aligned with contact points on the holder body. FIG. 19C shows the final assembly of the thermal cell holder for illustrative purposes. The holder lid is placed atop the holder body and affixed to the holder body using more than one screw. The depth of the deep and shallow pockets determines the amount of compression on the o-rings or other sealing means that surround the hole in the holder body and the deep pocket in the holder body. Adequate compression is required to form a tight seal around these features and prevent gases or liquids from escaping around the devices. Attaching the holder lid pushes down on the device stack, compressing these o-rings or other sealing means and forming a seal. Adequate compression force the gold contact pads on the thermal device on to the contact areas on the window device, and will also force the electrodes on the window device on to the underlying contact points, forming electrical contacts. The distance between the devices is set by the thickness of the contact pads on the thermal device and this distance does not change when the devices are pushed together when sealing the cell. A liquid cell holder comprising two window devices will have the same features as the electrochemical cell holder and the thermal cell holder with the exception that there can be one deep pocket in the holder of FIG. 7 because the length and width of the two window devices are substantially identical. Moreover, there is no need for electrical contacts. It should be appreciated that one or both of the window devices may have spacer material thereon. An example of a liquid cell holder can be seen in FIGS. 10 and 11, which will be described at length hereinbelow. In practice, liquids or gases can be flowed in and out of the liquid, electrochemical or thermal cells described herein through the supply lines without leaking to the outside environment. Electrical current and voltage can be supplied to the electrical or thermal device through the electrical supply lines. The holder can be placed in a TEM, the liquid, electrical or thermal device can be set to the desired current/voltage, and the type of liquid/liquids/gas/gases can be set applied to the sample using the supply lines. During imaging, the electron beam passes through the hole in the holder lid, strikes the sample on the heating membrane of the upper (window, thermal or electrical) device, passes through the window on the lower (window) device, then exits the gas cell through the hole on the bottom of the holder body. As discussed hereinabove, alignment of the two devices is critical. The membrane region of the window and electrical or thermal devices is preferably narrow so that the windows do not substantially distend when differential pressure is applied (atmospheric pressure inside the cell, vacuum outside the cell). Accordingly, the windows must be well-aligned since the electron beam can pass through the cell only where the windows are aligned. Narrow windows offer less area and therefore provide less margin for error with respect to alignment, so features that improve chip alignment are useful. Towards that end, another aspect of the present invention is a cell holder having a pocket(s) having contact points rather than straight edge walls so as to improve alignment of the devices in the cell holders. Referring to FIG. 10(d), an example of said pocket in a liquid cell holder body is illustrated. FIG. 10(a) shows a top view of the holder of FIG. 10(d) wherein the pocket has two contact points for each wall of the device(s), as shown more clearly in FIG. 12, where the hatched rectangle is intended to correspond to a device, e.g., window device, as described herein, having four straight edges. Having two contact points for each edge of the device reduces the likelihood that debris in the pocket can impact the device alignment. When the pocket accommodates two equally sized devices (e.g., for the liquid cell), the vertical contact points extend the full depth of the cavity, so the two chips see the same contact points and are therefore aligned to each other. It should be appreciated that the liquid cell holder of FIG. 10(a) and (d) has two contact points for each edge, however, the pocket can have at least one straight edge so long as at least one edge includes the contact points illustrated in FIG. 10(a) and (d). Another embodiment of the pocket is shown in FIG. 13, wherein the pocket accommodates two different sized devices (e.g., a window device with an electrical or a thermal device). In the embodiment of FIG. 13, both devices, having the same width, share width contact points on opposite edges while the length contact points are unique for each device. It should be appreciated that the contact points can be tooled to be any shape (e.g., hemispherical, square, triangular, etc.) or size as readily determinable by the skilled artisan. The liquid cell holder body of FIG. 10 correspondingly mates with the liquid cell holder lid of FIG. 11 and the two together make up the sample holder, e.g., as illustrated in FIGS. 5-7. The sample holder can be attached to the barrel (200) of FIG. 1. Moreover, as illustrated in FIGS. 10(e) and 10(f), multiple input/output ports are illustrated. It should be appreciated that only one input and only one output hole is contemplated. Further, the electrochemical cell holder and the thermal cell holders described herein (e.g., FIGS. 7, 8A-8C and 9A-9C) which are shown as having straight edge pockets preferably include at least one edge having the contact points of FIG. 10(a), 10(d), or 13, and most preferably, all of the edges include contact points. In another aspect, the use of multiple inputs/outputs to the sample holder in order to introduce multiple reagents during use is described. This allows the user to image chemicals as they mix/react in real time within the cell. It also improves time resolution since a chemical can be loaded, then released at a precise moment into the cell. For example, if a live cell is being imaged, the user can watch the live cell in flowing liquid, then introduce a fixative through a second input to instantaneously fix the cell. Alternatively, two different liquids can be input from either side of the liquid cell, react in the cell (between the windows), then released from a common output. A schematic of the multiple input/output teaching is illustrated in FIG. 14. It should be appreciated that the liquid cell, thermal cell, or electrochemical cell described herein can have one input and one output or any combination of multiple inputs/outputs as readily determined by the skilled artisan. The cell holders and lids described herein are preferably titanium or brass and are died to guarantee very vertical and parallel pocket edges. In practice, the first device (e.g., a window device, a heating device or an electrical biasing device) is loaded into the holder body, followed by the place of a sample on the membrane of the first device, the placement of a second device (e.g., a window device, a heating device or an electrical biasing device) in the holder body, and the placement of the holder lid over the holder body comprising the first and second devices and the sample. The holder lid can be secured to the holder body using at least one screw or other fastening means. The electrochemical cell holder is inserted into the electron microscope and the sample can be imaged in a liquid or gas environment using the holder described herein. In another aspect, a method of imaging a sample in a liquid and/or gaseous environment in an electron microscope is described, said method comprising inserting a sample in a sample holder, inserting the sample holder comprising the sample in an electron microscope, introducing a liquid and/or gas to the sample in the sample holder, and imaging the sample in the liquid and/or gaseous environment, wherein the sample holder comprises a sample holder body and a sample holder lid, wherein the sample holder body comprises at least one pocket having a pocket bottom and pocket walls for the positioning of two microelectronic devices therein, and the sample lid has a top side and a bottom side, as described herein. It should be appreciated that the two microelectronic devices may be the same as or different from one another and can comprise a device selected from the group consisting of a window device, a heating device, a electrical biasing device, and combinations thereof. Although the invention has been variously disclosed herein with reference to illustrative embodiments and features, it will be appreciated that the embodiments and features described hereinabove are not intended to limit the invention, and that other variations, modifications and other embodiments will suggest themselves to those of ordinary skill in the art, based on the disclosure herein. The invention therefore is to be broadly construed, as encompassing all such variations, modifications and alternative embodiments within the spirit and scope of the claims hereafter set forth.
053902289	abstract	A temperature distribution of an object, such as an optical element, onto which radiation energy is irradiated, is measured. The change of the shape of the object is controlled by varying the temperature of a part of the object on the basis of the measured temperature distribution to stabilize the shape of the object. Also, the shape of the object being irradiated is stabilized by causing the same temperature distribution in the object when in the thermally stable condition to be generated in the object while it is being irradiated. If the shapes of masks used to manufacture semiconductor devices are stabilized by using the above methods, highly integrated semiconductor devices can be manufactured.
	summary
	description	1. Field of the Invention The present invention relates to a focused ion beam (FIB) system for processing a specimen by directing a focused ion beam at the specimen and, more particularly, to a FIB system which calculates the processing conditions while the operator is monitoring the specimen after optical conditions are selected. 2. Description of Related Art A FIB system is a tool for processing a specimen by sharply focusing an ion beam produced from an ion source and directing the beam at the specimen so as to etch it. Among applications of such FIB systems, etching techniques relying on FIB have become widely spread. FIB systems using these techniques are widely used in defect analysis of semiconductor devices and specimen preparation in transmission electron microscopy as well as in micromachining. Especially, in three-dimensional analysis of semiconductor devices that has attracted the greatest attention, FIB systems are becoming indispensable tools. FIG. 1 shows the structure of a FIB system. The inside of the body 1 of the system is evacuated. The body 1 has a specimen chamber 1a in which a specimen stage 3 is placed. A specimen 2, such as a semiconductor device, is placed on the stage 3. Also contained in the body 1 are an ion source 5 for producing an ion beam 4, an extraction electrode 6 for extracting ions from the ion source 5, accelerating electrodes 7, condenser lenses 8 for focusing the ion beam, beam-blanking electrodes 9, multiple variable apertures 10, beam-deflecting electrodes 11 for scanning the ion beam in two dimensions, and an objective lens 12. A detector 13 for detecting secondary charged particles produced from the specimen 2 is also installed in the specimen chamber 1a. Electrostatic lenses are used for the condenser lenses 8 and objective lens 12. Some components (e.g., condenser lenses 8, multiple variable apertures 10, beam-deflecting electrodes 11, and objective lens 12) of the body 1 of the FIB system are driven by a FIB driver portion 14 that is under control of a computer 15. The computer 15 has an arithmetic unit 16, an input device 17, and a monitor 18 (e.g., a cathode-ray tube (CRT) or liquid crystal display (LCD)). The arithmetic unit 16 has RAM and HDD which are incorporated therein or attached thereto. For example, where the amount of current of the ion beam hitting the specimen 2 is varied, the FIB driver portion 14 controls the condenser lenses 8 and objective lens 12 to control the intensities of the lenses. This varies the degree of focusing of the beam. An appropriate aperture is selected from the multiple variable apertures 10 mounted in the optical path of the ion beam 4. In this way, the amount of the passing ion beam is controlled. Where the ion beam 4 is scanned over the specimen 2 in two dimensions or raster-scanned, a scan signal is supplied to the beam-deflecting electrodes 11 from the FIB driver portion 14. The specimen 2 is placed on the specimen stage 3. The stage 3 is designed to be capable of being moved in two dimensions within a horizontal plane, rotated, and tilted by a stage control portion 19, which is under control of computer 15. Ions are extracted from the ion source 5 by the extraction electrode 6. The ions are accelerated by the accelerating electrodes 7. The ion beam 4 of the accelerated ions is sharply focused onto the specimen 2 by the condenser lenses 8 and objective lens 12. The beam position on the specimen 2 is scanned by supplying the scan signal to the beam-deflecting electrodes 11. As a result, a desired portion of the specimen is cut or processed by the ion beam. The intensity of the ion beam 4 is controlled by the computer 15 via the FIB driver portion 14 such that the specimen 2 is not processed. The beam is scanned over the specimen 2 in two dimensions. Secondary electrons emanating from the specimen 2 are detected by the secondary electron detector 13. Image processing is performed by the arithmetic unit 16 of the computer 15 and then a secondary electron image is displayed on the monitor 18. Today, dual-beam systems each consisting of a conventional in-line scanning electron microscope (SEM) to which FIB capabilities are added have also become widespread. The dual-beam systems are described, for example, in Japanese Patent Laid-Open No. H7-37538. A dual-beam system (FIB/SEM instrument) is a combined instrument capable of playing the role of the conventional FIB instrument that etches a specimen as a semiconductor defect analysis tool and then moves the specimen onto a SEM to observe the specimen. This combined instrument has the advantage that it can perform SEM imaging similarly to an ordinary, single-function FIB machine. That is, an ion beam is directed at the top surface of a specimen. A desired portion is etched. After completion of the etching, the etched cross section can be immediately observed as an SEM image without moving the specimen. As a result, the combined instrument exhibits excellent capabilities in defect analysis and shortens the process sequence time. Concomitantly, the yield management can be done at an improved rate. Furthermore, the combined instrument has a small footprint because of the combined capabilities. The cost can also be reduced. The above-described FIB/SEM instrument roughly consists of a FIB control portion, a SEM control portion, and a stage control portion for controlling a specimen stage. These portions are controlled by a computer. The FIB instrument etches a specimen by directing an ion beam at the specimen such that the beam impinges on the specimen normally from vertically above it under the control of the computer. In the SEM, an electron beam impinges on the cross section of the formed hole at an angle of 30° with respect to the specimen surface to permit observation of the cross-sectional morphology. Where milling is done by a FIB system, it has been heretofore necessary to manually set parameters or select an appropriate setting file and utilize it. The parameters include (1) the size of the processed region, (2) the intensity of the used ion beam, (3) the depth of the cut hole and the kind of the specimen or the dose of the illuminating ion beam, and (4) processing and scanning conditions (dwell time (DT) per hit point and the dwell point spacing (DPS)). However, manual setting of the parameters (1)-(4) above or selection of an appropriate setting file depends on the knowledge and experience of each individual operator of the FIB system. Therefore, much labor and time are required to set the parameters. Furthermore, if different operators set different parameters in processing the same material, different processing results will arise. In addition, if any set parameter is disabled because of the hardware limitation, the settings are invalidated. It is an object of the present invention to provide a focused ion beam (FIB) system capable of automatically setting processing and scanning conditions (processing/scanning conditions) when a specimen is processed. A focused ion beam system associated with an embodiment of the present invention solves the foregoing problems and processes a specimen by directing a focused ion beam at the specimen. This system comprises an ion beam source for producing the ion beam, condenser lenses for focusing the produced ion beam, multiple variable apertures for selectively limiting the electrical current of the ion beam focused by the condenser lenses, a deflection portion for deflecting the focused ion beam whose current has been selectively limited by the apertures, an objective lens for focusing the deflected ion beam onto the specimen at a desired location, a specimen stage for moving the specimen, an input portion for accepting data entered by a human operator, a control portion, a setting condition data output portion, and a driver portion. The control portion selects optical conditions for the condenser lenses, multiple variable apertures, deflection portion, and objective lens based on the data entered into the input portion. The control portion also automatically calculates processing and scanning conditions of the focused beam on the specimen according to the selected optical conditions. The setting condition data output portion outputs data based on the optical conditions and the processing and scanning conditions selected and calculated by the control portion. The driver portion drives the condenser lenses, variable apertures, deflection portion, and objective lens based on the optical conditions and processing/scanning conditions outputted from the setting condition data output portion. In this FIB system, the control portion preferably calculates the processing and scanning conditions automatically according to the size of the observed region on the specimen entered into the input portion. Furthermore, the control portion preferably sets the dwell time of the ion beam on one point on the specimen and the dwell point spacing automatically as the processing and scanning conditions. Furthermore, in this FIB system, the control portion preferably calculates the diameter of the ion beam used for processing according to the size of the processed region of the specimen entered into the input portion, and selects an optical condition file matched to the diameter of the beam. In addition, the control portion preferably selects the optical condition file automatically based on the calculated diameter of the beam, the file defining a mode of operation in which the condenser lenses, objective lens, multiple variable apertures, and deflection portion are driven by the driver portion. The FIB system, according to the present invention, automatically calculates and sets the diameter of the ion beam used for processing according to the “size of the processed region” entered by the operator in this way. Furthermore, the system automatically calculates and sets the diameter of the ion beam used for processing according to the “degree of finish” of the processed specimen. In addition, the system automatically calculates and sets beam processing and scanning conditions (i.e., (1) dwell time (DT) per hit point and (2) dwell point spacing (DPS)) according to the diameter of the used ion beam. Additionally, the system automatically calculates and sets the beam processing and scanning conditions (i.e., (1) dwell time (DT) per hit point and (2) dwell point spacing (DPS)) according to the depth of the processed region and dose. Of course, these functions may be appropriately combined. The FIB system associated with the present invention can automatically set processing and scanning conditions under which a specimen is processed. That is, beam diameter, dwell time per hit point, and dwell point spacing used for the processing are calculated and automatically set according to the size of the processed region and the degree of finish that the operator wants. Furthermore, the dwell point spacing used when the ion beam is scanned is automatically calculated and set optimally according to the used ion beam. Consequently, it is possible to prevent such a situation that a dwell point spacing that is too wide for the beam diameter is set; otherwise, the processed surface would be made discontinuous. Additionally, the dwell time per hit point and dwell point spacing which reduce the number of frames used for processing down to a bare minimum are automatically calculated and set optimally. Consequently, damage to the specimen caused by blanking tail can be minimized. Further, deterioration of the finished shape can be prevented because the dwell time per hit point and dwell point spacing are automatically calculated and set optimally. Other objects and features of the invention will appear in the course of the description thereof, which follows. The best mode for carrying out the present invention is hereinafter described. This mode is a focused ion beam (FIB) system that can be widely applied to defect analysis of semiconductor devices and specimen preparation in transmission electron microscopy (TEM), as well as to micromachining. FIG. 2 shows the structure of the FIB system. This system is similar to the prior art FIB system already described in connection with FIG. 1 except that a computer 15 selects optical conditions (described later) according to data entered by the operator indicating the size of the processed region and the degree of finish of the processed material and that the computer automatically sets processing and scanning conditions (described later) according to the size of the observed region indicated by data entered by the operator. The resulting data are supplied to a setting condition data output portion 20. The body 1 of the FIB system has an ion beam source 5, an extraction electrode 6, accelerating electrodes 7, condenser lenses 8, beam-blanking electrodes 9, multiple variable apertures 10, beam-deflecting electrodes 11, and objective lens 12. A specimen 2 is placed on a specimen stage 3 within a specimen chamber 1a. The FIB system further includes the computer 15 made up of an arithmetic unit 16, an input device 17 connected with the arithmetic unit 16, and a monitor 18 connected with the arithmetic unit. The FIB system further has the FIB driver portion 14 for driving given internal parts of the body 1 in accordance with the set condition data from the setting condition data output portion 20 that is mounted between the arithmetic unit 16 and the FIB driver portion 14. The FIB system is further equipped with a stage control portion 19 for controlling the specimen stage 3. Inside the body 1 of the FIB system, the ion beam source 5 produces an ion beam 4. The extraction electrode 6 extracts ions from the ion source 5. Acceleration electrodes 7 accelerate the ions, which have been extracted from the ion source 5 by the extraction electrode 6. The condenser lenses 8 focus the ion beam produced by the ion beam source 5. The beam-blanking electrodes 9 turn on and off the impingement of the ion beam 4 focused on the specimen 2 by the condenser lenses 8. The multiple variable apertures 10 selectively limit the current of the ion beam that is made to impinge on the specimen by the blanking electrodes 9. The beam-deflecting electrodes 11 deflect the focused ion beam 4 whose current has been selectively limited by the multiple variable apertures 10. The objective lens 12 focuses the ion beam 4, which has been deflected by the deflecting electrodes 11, at a given position on the specimen 2. The specimen stage 3 moves the specimen 2 in two dimensions within a horizontal plane, rotates the specimen, or tilts it. The input device 17 accepts data entered by the operator. The arithmetic unit 16 selects optical conditions for the condenser lenses 8, variable apertures 10, beam deflecting electrodes 11, and objective lens 12 based on the data entered into the input device 17. The arithmetic unit 16 also automatically calculates the processing and scanning conditions under which the ion beam 4 is scanned to process the specimen 2, according to the selected optical conditions. The setting condition data output portion 20 outputs data based on the optical and processing/scanning conditions selected and calculated by the arithmetic unit 16. The FIB driver portion 14 drives the condenser lenses 8, blanking electrodes 9, apertures 10, deflecting electrodes 11, and objective lens 12 based on the optical and processing/scanning conditions outputted from the setting condition data output portion 20. FIG. 3 illustrates the manner in which data processing and control are performed by the computer 15, setting condition data output portion 20, and FIB driver portion 14. Plural optical condition files have been previously registered in a storage portion (not shown) of the arithmetic unit 16 of the computer 15. FIG. 4 is a table illustrating the optical condition files in which sets of control data about the lenses, deflection system, and multiple movable apertures necessary to obtain various ion beam diameters are stored in a corresponding manner to file numbers. Optical condition file numbers “beam 1” to “beam 10” are assigned in the order of increasing or reducing the beam diameter. With the optical file condition number “beam 1”, the lenses, apertures, and deflection system are set to obtain a beam diameter of φ1. With optical condition file numbers “beam 2” to “beam 10”, the lenses and so on are set to have beam diameters of φ2 to φ10. In FIG. 3, if the operator enters the size of the processed region (1) as input data D-IN into the input device 17, the arithmetic unit 16 of the computer 15 calculates the used beam diameter from the input data. The “size of the processed region” is the size of a region on the specimen that the operator wants to have by means of processing using the ion beam. The arithmetic unit 16 executes a program for finding the beam diameter from the processed region size specified by the operator. The beam diameter is calculated by the calculational program based on the ratio to the side (width) of the rectangle of the size of the processed region entered by the operator. For example, if the width of the rectangle is 10 μm, the ratio is 1/10. The beam diameter is calculated to be 1 μm. An “optical condition file” having a beam diameter closest to the calculated beam diameter is automatically selected as a reference file. Then, the operator enters data about the “degree of finish” as input data D-IN from the input device 17. “Optical condition files” having beam diameters approximate to (slightly greater and smaller than) the beam diameter assumed in the reference file are selected. The “degree of finish” is a setting regarding the niceness and accuracy of ion beam processing that the operator wants. This is a qualitative set item including shear droop at edges caused during cutting operation. For example, where a file of “beam 3” shown in FIG. 4 is selected as the reference file, files placed around the file of “beam 3” are selected based on the data about the degree of finish. That is, any one of files of “beam 2” and “beam 4” is selected based on the data about the degree of finish. In particular, the beam diameter is determined as a reference diameter from the size of the region. Beam diameters which are respectively slightly larger and smaller than the reference diameter are selected according to the degree of finish selected by the operator. Optical conditions are necessary which are used (i) to determine the current values supplied to the lenses including the condenser lenses 8 and objective lens 12 of the electrostatic type in order to make the focused ion beam have the calculated beam diameter, (ii) to determine which of the multiple variable apertures 10 is selected, (iii) or to determine the current supplied to the deflecting electrodes 11. Therefore, an optical condition file is automatically selected by the computer 15. The selection depends on the beam diameter. As the beam diameter is increased, the beam intensity is also increased. Therefore, where a focused ion beam having a larger diameter is used (i.e., an intenser beam is used), the specimen can be processed in a shorter time than where a focused ion beam having a smaller diameter is used. In this way, sets of optical condition files are selected according to entered information about the size of the processed region and about the degree of finish. The arithmetic unit 16 calculates the used beam diameter based on input conditions about the “size” of the processed region. An optical condition file having a beam diameter closest to the result of calculation is selected as a reference. Furthermore, optical condition files having beam diameters respectively slightly greater and smaller than the reference beam diameter are selected based on the settings on the “degree of finish”. The processing and scanning conditions define the dwell time (DT) per hit point when the specimen 2 is irradiated with a focused ion beam, the dwell point spacing (DPS), and the number of frames. The processing and scanning conditions calculated by the arithmetic unit 16 are set using information (i.e., the size of the processed region, cut depth+specimen kind, or dose) entered by the operator. In the present embodiment, the processing and scanning conditions are calculated as follows. In fundamental scanning, the dwell point spacing is based on a frame of microscope image accepted for setting of a processed region. The dwell point spacing obtained where the resolution of the microscope image is 2,560×1,920 pixels is used as default dwell point spacing for processing. For example, a dwell time per hit point of 1 μs is used as a default dwell time. The default processing and scanning conditions set as described above are modified using conditional formulas (described later). Thus, actually used processing and scanning conditions are calculated. Accordingly, in the present method, the processing and scanning conditions are fundamentally different among different processed regions. The arithmetic unit 16 calculates the processing and scanning conditions. After selecting the above-described optical condition file, the operator enters data D-IN from the input device 17 as shown in FIG. 3. The entered data include (1) the size of the processed region, (2) cut depth and specimen kind, or (3) dose. The arithmetic unit 16 performs calculations using a calculational program P-CPU and IOS data D-IOS offered by the optical condition file. The “cut depth+specimen kind” or “dose” is an input item determining the cut depth that the operator wants. Thus, the amount of electric charge (dose×area of the processed region) implanted into the processed region is computed. The dose is the number of ion beams implanted to a given area. The size of the processed region is a region that is scanned by the ion beam in two dimensions. This is associated with the magnification of the observed microscope image. In this embodiment, the microscope image is made up of 2,560×1,920 pixels. The size of the processed region is specified by the operator. FIG. 5 illustrates the ratio of the width D of the displayed microscope image to the scanning width L of the focused ion beam on the specimen 2 corresponding to the width D, the image being formed by detecting secondary electrons emanating from the specimen 2 when the ion beam is scanned over the specimen 2 in two dimensions. The ratio is referred to as the scanning magnification or simply as the magnification. The dwell time (DT) and the dwell point spacing (DPS) that are the processing-and-scanning conditions are determined as default conditions, depending on the magnification of the microscope image of the specimen viewed by the operator. That is, DT and DPS are determined according to the size of the observed region. The calculational program P-CPU consists of plural calculational formulas. Optimum setting values for processing conditions corresponding to the input data D-IN specifying the observed region are calculated for each processed region. The dose was found at the previous step. Therefore, the amount of ions implanted into a unit area can be computed. Then, the processing and scanning conditions can be automatically calculated from the computed amount. In the case of an unknown specimen for which the processing rate is not set, the dose may be directly entered instead of the depth and specimen as described later. This method is different from the method adopted in the prior art product (i.e., set values judged to be appropriate according to input data are selected from a finite number of preset data sets). The output data indicative of the results of calculations performed by the arithmetic unit 16 are used for control of processing and scanning and for IOS control. IOS data are held in the aforementioned plural (e.g., ten) optical condition files, and are previously stored in the arithmetic unit 16 as data for adjusting the system. Referring to FIG. 3, processing and scanning conditions (output data 1) D-OUT1 calculated by the arithmetic unit 16 using the calculational program P-CPU are supplied to the processing and scanning control portion 14-1 of the FIB driver portion 14 by the setting condition data output portion 20. The processing and scanning control portion 14-1 drives the condenser lenses 8, beam-blanking electrodes 9, multiple variable apertures 10, beam-deflecting electrodes 11, and objective lens 12 inside the body 1 of the FIB system to obtain the dwell time (DT) per hit point, dwell point spacing (DPS), and the number of frames which are stipulated in the processing condition file (output data 1) D-OUT1. The reference file selected from the optical condition files previously held in the arithmetic unit 16 is outputted as output data 2D-OUT2 to the setting condition data output portion 20 and supplied to the optical control portion 14-2 of the FIB driver portion 14. The optical control portion 14-2 sets the optical lens system, apertures, and deflection system to obtain a desired beam diameter based on the output data 2D-OUT2. FIG. 6 is a flowchart illustrating selection of the optical condition file performed by the computer 15 and a subsequent sequence of operations for calculating the processing and scanning conditions. First, the operator is prompted to enter a processed region size from the input device 17 (step S1). Then, the operator is prompted to enter a degree of finish (step S2). The arithmetic unit 16 calculates the ion beam diameter according to the entered size of the processed region and selects an optical condition file in which the calculated ion beam diameter is assumed, as a reference file from previously stored ten optical condition files. In step S2, a degree of finish indicating high speed or high accuracy has been entered and so the arithmetic unit 16 selects files adjacent to the reference file according to the degree of finish (step S3). In step S4, the operator is prompted to enter a kind of specimen 2. Since the processing speed is different according to different kinds of specimen, entry of the kind of the specimen is required. The processing time differs according to the specified depth and specimen kind. The processing speed determining the cut depth is preset by the operator according to the specimen kind. Therefore, a decision is made as to whether the processing speed for the kind of specimen entered in step S4 has been registered (step S5). If the decision is affirmative (Yes), control goes to step S6, where a cut depth is automatically entered. Then, the dose is automatically calculated according to the depth (step S7). Then, DT and DPS are calculated automatically as the processing and scanning conditions (step S8). If the decision at step S5 is that the processing rate for the specimen entered at step S4 is not registered, the operator is prompted to enter the dose manually (step S9). A sequence of operations for correction performed after the calculation of the processing and scanning conditions and its purpose are next described by referring to Table 1 and FIG. 7. In FIG. 7, the sequence of operations for correction is shown as steps S11 to S14 in the order of priority given to correction. TABLE 1PurposeSolutionTo form clean shape by processinga) (processing DPS) ≦ 0.25(beam diameter)b) 20 ≦ FNTo increase the speed at whichc) FT ≦ 1 sreal-time monitor screen is updatedTo reduce blanking taild) 0.25 s ≦ FT First, the DPS (dwell point spacing) of the processing and scanning conditions is kept to within one-fourth of the beam diameter (step S11). In other words, more than ¾ of the beam diameter overlaps. This prevents undulations of the shapes formed by processing, as well as formation of porous cavities. Then, the number of frames FN of scanning processing is kept above 20 (step S12) to prevent distortion of the shape obtained by processing and prevent the cut depth from ramping. Then, the number of frames FT is kept to less than 1 second such that the image on the real-time monitor is constantly updated quickly (step S13). To make the frame time (FT) have a value of more than 0.25 second, the number of frames is suppressed to prevent the number from becoming excessively large (step S14). This suppresses the effects of the blanking tail during scanning for processing. As described thus far, the FIB system of the structure shown in FIG. 2 automatically calculates and sets the ion beam diameter used for processing, according to the “size of the processed region” entered by the operator. Furthermore, the system automatically calculates and sets the ion beam diameter used for processing according to the “degree of finish”. In addition, the beam processing and scanning conditions (i.e., (1) dwell time (DT) per hit point and (2) dwell point spacing (DPS)) are automatically calculated and set according to the used ion beam diameter. Furthermore, the beam processing and scanning conditions (i.e., (1) dwell time (DT) per hit point and (2) dwell point spacing (DPS)) are automatically calculated and set according to the depth of the processed region and the dose. Of course, these functions can be combined appropriately. Accordingly, the FIB system of the structure shown in FIG. 2 yields the following advantages. (1) The controllability can be improved by automated setting of the processing method using the scanned ion beam. In the past, the operator has been urged to select a used beam intensity according to the size of the processed region and the desired degree of finish, relying on his own knowledge and experience. However, according to the FIB system of the present embodiment, the beam diameter used for processing, dwell time per hit point, and dwell point spacing are calculated and automatically set according to the size of the processed region and the degree of finish that the operator wants. (2) Discontinuous processing can be suppressed. In the FIB system that shoots focused ion beams, ion beams having different intensities have different beam diameters. Accordingly, in the past, the operator has had to set the dwell point spacing taking account of the ion beam diameter. According to the FIB system of the present invention, however, the dwell point spacing is automatically calculated and optimally set when the ion beam is scanned according to the used ion beam. Consequently, the dwell point spacing can be prevented from becoming too great relative to the beam diameter; otherwise, a discontinuously processed surface would be produced. (3) Furthermore, specimen damage due to blanking tail can be suppressed. During FIB micromachining, the specimen surface is damaged by the ion beam irradiation. When the ion beam is scanned, the blanking function is necessary. That is, the beam is deflected greatly at the end of each raster line and impingement of the beam on the specimen surface is stopped. However, when blanking is carried out, the ion beam passes through specimen surface portions located outside the processed region, though in a short time. This produces damage to the specimen surface. This damage is known as blanking tail. To minimize the damage due to blanking tail, it is necessary to minimize the number of frames scanned during processing. Accordingly, in the past, the operator has had to set the dwell point spacing taking account of the number of frames. On the other hand, in the FIB system of the present embodiment, the dwell time per hit point and the dwell point spacing are automatically calculated and optimally set such that the number of frames is reduced to a bare minimum. Consequently, damage to the specimen due to blanking tail can be suppressed to a minimum. (4) The finishing accuracy can be improved. In FIB processing utilizing scanning of a focused ion beam, if the scan speed of the beam decreases below a threshold value, the finished shape is disfigured severely. Accordingly, in the past, the operator has had to set the dwell time per hit point and the dwell point spacing taking account of the scan speed of the beam. On the other hand, according to the FIB system of the present embodiment, the dwell time per hit point and the dwell point spacing are automatically calculated and set optimally and, therefore, deterioration of the finished shape can be prevented. Another embodiment of the present invention is next described. This embodiment provides a dual-beam (DB) system in which a FIB system and a scanning electron microscope (SEM) are combined. FIG. 8 shows the structure of the dual-beam system. This system has an ion beam column 30 having the same structure as the FIB system already described in connection with FIG. 2. The column 30 produces an ion beam that is directed at a certain region of a specimen 2 to process the region. The processed region is scanned with an electron beam produced by an electron beam irradiation column 40 consisting of the SEM. Secondary electrons are produced in response to the electron beam scanning, and are detected by a secondary electron detector 13. Thus, a scanned image is obtained. Included within the electron beam irradiation column 40 are an electron source 41, extraction electrodes 42 for extracting ions from the electron source 41, an anode 43, condenser lenses 44, beam-blanking electrodes 45, apertures 46, beam-deflecting electrodes 47, and an objective lens 48. The electron source 41, condenser lenses 44, beam-blanking electrodes 45, apertures 46, beam-deflecting electrodes 47, and objective lens 48 are controlled by a SEM control portion 50. For example, where the current of the electron beam hitting the specimen 2 is varied, the amount of electron beam produced from the electron source 41 is controlled. In order to vary the beam diameter on the specimen 2, the intensities of the condenser lenses 44 and objective lens 48, for example, are controlled so as to control the degree of focusing of the electron beam. The impingement of the ion beam on the specimen 2 is turned on and off by controlling the beam-blanking electrodes 45. The electron beam scanning of the surface of the specimen 2 is controlled by the beam-deflecting electrodes 47. The SEM control portion 50 is controlled by the computer 15 similarly to the FIB driver portion 14. In this dual-beam (DB) system, the specimen 2 is processed by the ion beam column 30 consisting, for example, of a FIB system. This processing is performed by producing an ion beam from the ion source 5, sharply focusing the beam 4 onto the specimen 2 by the condenser lenses 8 and objective lens 12, and raster-scanning the beam over the specimen by the beam-deflecting electrodes 11. At the same time, the specimen stage 3 is moved in a direction perpendicular to the raster lines. As a result of the scanning and movement, a hole is formed in a desired region of the surface of the specimen 2 by the ion beam processing. Then, the electron beam is directed at the cross section of the hole from the electron beam irradiation column 40 consisting of the SEM to scan the cross section in two dimensions. Secondary electrons produced in response to the scanning are detected by the secondary electron detector 13. The output signal from the detector 13 is supplied to the monitor 18 via the arithmetic unit 16 and so a SEM image of the cross section of the processed portion is obtained. Of course, the ion beam column 30 included in the dual-beam system and consisting of the FIB system operates on the principle already described in connection with FIGS. 3-5. That is, the diameter of the ion beam used for processing is automatically calculated and set according to the “size of the processed region” entered by the operator. Furthermore, the diameter of the ion beam used for processing is automatically calculated and set according to the “degree of finish”. In addition, the dual-beam system automatically calculates and sets beam processing and scanning conditions (i.e., (1) dwell time (DT) per hit point and (2) dwell point spacing (DPS)) according to the diameter of the used ion beam. Additionally, the system automatically calculates and sets the beam processing and scanning conditions (i.e., (1) dwell time (DT) per hit point and (2) dwell point spacing (DPS)) according to the depth of the processed region and dose. Of course, these functions may be appropriately combined. Hence, the system yields the same advantages as the above-described system. Having thus described my invention with the detail and particularity required by the Patent Laws, what is desired protected by Letters Patent is set forth in the following claims.
	summary
045405450	abstract	A device for securing a fuel assembly to a nuclear reactor pressure vessel comprises upper and lower tie plates for supporting the fuel assembly at its lower and upper ends, spacers for supporting the fuel assembly at intermediate portions between the both tie plates, a channel box surrounding the fuel assembly, and a stationary handle attached to the upper tie plate. A movable handle is attached to the stationary handle to be vertically slidable with respect thereto and the movable handle is always urged downwardly by a coil spring. Hook members are connected to the stationary handle to be rotatable in accordance with vertical movement of the movable handle such that the lower end of the hook member engages with an upper grid of the pressure vessel when the movable handle is lowered to a low portion and disengages therefrom when the movable handle is lifted to an upper position.
041475887	description	Referring now to the attached drawings, the proposed recharging device for a fast-neutron reactor, shown in FIGS. 1 and 2, comprises a recharging mechanism 1 installed in rotating plugs 2 which are rotatable around their axes in a nuclear reactor vessel 3 above a reactor core 4. The device further includes a mechanism for transferring fuel assemblies 5 and rods 6 of the control and safety system. Elsewhere in this text, when mention is made of the fuel assemblies 5 and rods 6 in conjunction, these will be referred to as "assemblies" for reasons of brevity. The assembly transfer mechanism comprises an elevator 7 with a charging socket 8 and a recharging socket 9, and an assembly handling mechanism 10. The proposed device also includes a storage drum 11 for new assemblies and a storage drum 12 for spent assemblies. The recharging mechanism 1 has a grip 13 which terminates in a stop 14. FIG. 3 is a sectional view taken on line III--III of FIG. 1 and shows compartments 15 in the storage drum 11, which are all identical and intended to receive assemblies. A rod 6 of the control and safety system is placed in a sleeve-type holder 16, so that its overall dimensions are equal to those of the fuel assembly 5. FIG. 4 is an elevation view of a rod 6 of the control and safety system, placed in the holder 16 which in its upper portion is provided with a collet 17 having a cone-shaped inner surface 18. The collet 17 interacts with an encircling groove 19 provided on the surface of the rod 6, near its profiled head 20. Between the bottom of the holder 16 and the end face of the rod 6, there is arranged a spring 21 with a pusher 22, intended to push the rod 6 from the holder 16 during a recharging operation. When placed in the holder 16, the rod 6 is referred to as an assembly 23. FIG. 5 is a view of the unit A of FIG. 1, taken at a moment when the new rod 6 is disengaged from the holder due to the interaction of the stop 14, protruding from the grip 13 of the mechanism 1, with the cone-shaped surface of the collet 17 of the holder 16. FIG. 6 is a view of the unit B of FIG. 1, taken at a moment when the spent rod 6 is engaged with the holder 16 due to the interaction of a grip body 24 of the mechanism 10 with the profiled head 20 of the rod 6. The proposed device for recharging a fast-neutron reactor operates as follows. Prior to the recharging, a new rod 6 (FIG. 4) of the control and safety system is placed in the sleeve-type holder 16; the collet 17 is received in the encircling groove 19, making up the assembly 23 which is inserted into any compartment 15 (FIG. 3) of the storage drum 11 (FIG. 2) for new assemblies, adjacent to the new fuel assemblies 5 placed in the similar compartments 15. The rotor of the storage drum 11 for new assemblies is set into rotation, and one of the compartments 15 (FIG. 3), holding the assembly 23, is matched with the axis of the mechanism 10 (FIGS. 1 and 2). The latter extracts the assembly 23, transfers it and places it in the charging socket 8 of the elevator 7. The latter transfers the assembly 23 received in the socket 8 to the periphery of the reactor core 4. By rotating the plugs 2, the mechanism 1 is matched with the axis of the socket 8 which holds the assembly 23. The cone-shaped surface of the stop 14 protruding from the grip 13 of the mechanism 1 slides over the cone-shaped surface 18 of the jaws of the collet 17, unclamps the latter and disengages it from the encircling groove 19 provided on the surface of the rod 6, near its profiled head 20. At the same time the pusher 22 pushes the rod 6 upwards under the action of the spring 21, until the collet 17 leaves the encircling groove 19. The grip 13 of the mechanism 1 grips the profiled head 20 on the outside and withdraws the new rod 6 from the holder 16. As this takes place, the holder 16 remains in the socket 8 of the elevator 7. After this, by rotating the plugs 2, the mechanism 1 with the new rod 6 clamped in the grip 13 is directed at the transfer socket (not shown) provided in the screen of the core 4, wherein there is temporarily placed the new rod 6. The rotating plugs 2 then point the mechanism 1 at the spent rod 6 in the core 4, which is to be replaced. The latter rod 6 is removed with the aid of the grip 13 and transferred to the holder 16 found at this time in the socket 8 of the elevator 7. From its temporary storage place in the socket in the screen of the core 4, the new rod 6 is placed, with the aid of the rotating plugs 2 and the mechanism 1, in the vacated socket of the core 4, wherefrom the spent rod 6 has been just removed. The elevator 7 then moves the socket 8 with the holder 16 holding the spent rod 6 down the inclined surface until it comes into contact with the mechanism 10. As the mechanism 10 reaches the holder 16 with the spent rod 6, its body 24 presses upon the head 20 of the rod 6, whereby the spring 21 is compressed and the rod 6 is forced into the holder 16 until the collet 17 engages with the encircling groove 19. As a result, another assembly 23 is produced and placed by the mechanism 10 in the compartment 15 of the spent assembly storage drum 12. At this point the recharging process is completed. The recharging of fuel assemblies is carried out in a similar manner, with the exception of the operation of removing the new assembly from the holder and placing the spent assembly in this holder. However, the duration of both processes (the recharging of fuel assemblies and rods of the control and safety system) is equal, because the removal of a rod of the control and safety system from the holder and placing it in the holder are carried out simultaneously with the removal of this rod from or placing it in the respective socket of the core or compartment of the storage drum.
	abstract	A method of aligning a nuclear fuel bundle and handling selected fuel rods within the fuel bundle located in a spent fuel pool of a nuclear power plant. The bundle includes water rods, full-length and part-length fuel rods extending through a plurality of fuel spacers provided between top and bottom ends of the bundle, each spacer having a plurality of cells accommodating corresponding fuel and water rods. The method includes insertion of a rod grapple tool into a top end of a fuel bundle and down to a desired location within the bundle. A guide pin retrieval tool inserts into a side of the fuel bundle to remove a guide pin from a distal end of a gripper of the rod grapple tool, allowing the guide pin retrieval tool to grip a part-length fuel rod within the fuel bundle.
	description	A photon source in accordance with a feature of the invention operates in two stages to produce X-ray or extreme ultraviolet radiation. In a first stage, a central plasma is formed using multiple ion beams directed at a central plasma discharge region as described below. In a second stage, a heating current pulse is passed through the central plasma in order to heat and compress the plasma, raising its temperature and density. The astron source is a source of photons comprising a discharge chamber, a plurality of ion beam sources in the discharge chamber, each electrostatically accelerating a beam of ions of a working gas toward a plasma discharge region, and a neutralizing mechanism for at least partially neutralizing the ion beams before they enter the plasma discharge region. The neutralized beams enter the plasma discharge region and form a hot plasma that radiates photons. The astron principle that operates in the first stage of the photon source described above is illustrated in FIGS. 1A and 1B. The embodiment of the source shown in FIGS. 1A and 1B has a two gap ion acceleration structure 100. Acceleration structure 100 includes concentric spherical electrode shells 112, 113 and 114. The electrode shells 112, 113 and 114 have a plurality of sets of holes aligned along axes which pass through a central plasma discharge region 120. Thus, for example, holes 122, 123 and 124 in electrode shells 112, 113 and 114, respectively, are aligned along an axis 126 that passes through plasma discharge region 120. Each set of holes, such as holes 122, 123 and 124, defines an acceleration column 128. The spaces between electrode shells 112, 113 and 114 constitute acceleration gaps for electrostatic acceleration of ion beams. Thus, each acceleration column has two gaps in the embodiment of FIGS. 1A and 1B. The embodiment of FIGS. 1A and 1B includes 36 acceleration columns 128, arrayed in three sets of 12. Thus, the acceleration structure directs 36 ion beams toward plasma discharge region 120. However, different numbers of ion beams may be utilized within the scope of the invention. The electrode shells 112, 113 and 114 may be supported by insulating spacers 130. A plenum 132 having ports 134 encloses acceleration structure 100. A working gas is introduced, either in a pulsed mode or continuously, through ports 134 into a space 144 behind the outermost electrode shell 114. Some of the working gas flows down the acceleration columns 128. When the appropriate gas density is present in the acceleration columns, a pulsed voltage may be applied between electrode shells 112 and 114, with the polarity of electrode shell 114 being positive with respect to electrode shell 112. In the configuration of FIGS. 1A and 1B, provided the appropriate gas density is present and provided that sufficient voltage is applied, a pseudospark discharge develops simultaneously in each of the acceleration columns 128. The pseudospark discharge is characterized by the development of oppositely directed electron and ion beams that can have extremely high intensity. The ion beam exits from the negative polarity end of the acceleration column 128 at electrode shell 112 and progresses toward the central plasma discharge region 120. By correct adjustment of the working gas density at an exit region 146 of each of acceleration columns 128, most of the ions can be neutralized by resonant charge exchange, so as to form a neutral beam that propagates without deflection to the plasma in plasma discharge region 120. Those ions that are not neutralized contribute excess positive charge to each of the ion beams, causing electrons to be attracted from the nearby surface of electrode shell 112, which is already primed as a cathode due to the breakdown into a pseudospark discharge. Thus, the neutral atoms are accompanied by a nearly charge-balanced beam plasma, including the remaining unneutralized ions and electrons. The slow ions resulting from resonant charge exchange define tracks that are favored for a conduction of a high current heating pulse in the second stage of device operation, as described below. Additional details and embodiments of the astron photon source are described in the aforementioned application Ser. No. 09/815,633, which is hereby incorporated by reference. In the second stage of device operation, the newly formed plasma is heated and compressed, or pinched, by passage through the plasma of a pulse of electric current. A first embodiment of a photon source which incorporates both the first stage of operation, wherein beams of ions are electrostatically accelerated toward a plasma discharge region and are at least partially neutralized, and the second stage, wherein an electric current is passed through the plasma discharge region, is shown in FIGS. 2A and 2B. FIG. 2A is a simplified cross-sectional side view of the photon source, and FIG. 2B is a cross-section defined by revolution of line Axe2x80x94A in FIG. 2A around axis 200. In FIGS. 2A and 2B, a central cathode shell, corresponding to electrode shell 112 in FIG. 1A, is divided into two half shells 202 and 204 that are electrically connected to a pulse voltage source 205. The anode shell of the photon source is divided into two half shells 212 and 214 which are electrically connected by a conductor 216. The working gas is introduced at low pressure through ports 218 and flows through passages 220 to enter hollow anode volumes 222 within anode half shells 212 and 214. Cathode half shells 202 and 204 are electrically isolated by insulator 225. The cathode half shells are electrically isolated from the respective anode half shells by insulators 227 and 229. A pulse voltage source 215 is connected between cathode half shells 202 and 204 and anode half shells 212 and 214. During the first phase of operation of the photon source shown in FIGS. 2A and 2B, a pulsed voltage V1 from pulse voltage source 215 is applied between anode half shells 212 and 214 and cathode half shells 202 and 204. In the absence of any applied voltage V2 from pulse source 205, the potential difference between cathode half shells 202 and 204 remains at zero. The combined cathode half shells are therefore pulsed negatively by voltage V1 relative to the combined anode half shells, and a discharge develops as described above in connection with FIGS. 1A and 1B. Neutralized beams from this discharge pass through a plasma discharge region 224 to form a small spherical plasma. At the same time, the passage of ions and energetic neutral atoms forms ionized tracks 230 between cathode half shell 202 and plasma discharge region 224, and ionized tracks 232 between cathode half shell 204 and plasma discharge region 224. The ionized tracks 230 and 232 lie on the surfaces of two cones that have their vertices located at plasma discharge region 224 and provide conducting paths between cathode half shells 202 and 204. During the second phase of operation, a pulsed voltage V2 from pulse voltage source 205 is applied between cathode half shells 202 and 204. The circuit is completed by conduction through the conical configuration of ionized tracks connecting cathode half shells 202 and 204. The current flows through the plasma in plasma discharge region 224, heating and compressing it via the magnetic pinch effect. The plasma temperature and density rise to the point where the desired X-ray or extreme ultraviolet radiation is emitted copiously. The radiation escapes the photon source in conical beams 234 that are relayed to the point of use by collecting optical surfaces (not shown in FIGS. 2A and 2B). The working gas pressure in the central part of the acceleration structure is preferably maintained in a range of about 1.0 to 100 millitorr. As noted above, one suitable working gas is xenon. Other suitable working gases include, but are not limited to, hydrogen, lithium, helium, nitrogen, oxygen, neon, argon and krypton. The ion beams may be pulsed or continuous, and the ion acceleration voltage V1 may be from 2 kV to 20 kV, but is not limited to this range. Voltage V1 may have a typical pulse duration of 100 nanoseconds to 10 microseconds, but may also be applied continuously. The heating voltage V2 is applied typically within 100 nanoseconds to 10 microseconds of the initial application of voltage V1 . The amplitude of voltage V2 is typically in the range of 100 volts to 10 kV, and the width of this pulse typically ranges from 10 nanoseconds to 1 microsecond. A second embodiment of a photon source in accordance with the invention is shown in FIGS. 3 and 4. FIG. 3 is a simplified cross-sectional side view of the photon source, and FIG. 4 is a cross-section defined by revolution of line Axe2x80x94A in FIG. 3 around axis 200. Like elements in FIGS. 2A, 2B, 3 and 4 have the same reference numerals. The embodiment of FIGS. 3 and 4 differs from the embodiment of FIGS. 2A and 2B by the addition of a transformer 211 for coupling pulsed electrical current from pulse voltage source 205 to cathode half shells 202 and 204. Transformer 211 includes multiple primary windings 208, a toroidal core 210, which may be of non-magnetic or magnetic material, and an armature or secondary 206, also having a toroidal configuration. Primary windings 208 are connected to pulse voltage source 205, and secondary 206, which may have a single turn, is connected between cathode half shells 202 and 204. During the first phase of operation, a pulsed voltage V1 is applied between the anode half shells 212 and 214 and secondary winding 206 that is connected to cathode half shells 202 and 204. In the absence of any applied voltage V2 from pulse voltage source 205 to primary windings 208, the electric potential between the cathode half shells 202 and 204 remains at zero. The combined cathode half shells are therefore pulsed negatively by voltage V1 relative to the combined anode half shells 212 and 214, and a discharge develops as described above. The neutralized beams of this discharge pass through plasma discharge region 224 to form a small spherical plasma. At the same time, the passage of ions and energetic neutral atoms forms ionized tracks 230 and 232 as described above. During the second phase of operation, a pulsed voltage V2 is applied simultaneously and in parallel across all the primary windings 208, with the result that a voltage is induced between cathode half shells 202 and 204 that are connected to opposite ends of transformer secondary 206. The transformer secondary circuit is completed by a conduction through the ionized tracks 230 and 232 connecting cathode half shells 202 and 204. The secondary current flows through the plasma in plasma discharge region 224, heating and compressing it via the magnetic pinch effect. As described above, the plasma temperature and density rise to the point where the desired X-ray or extreme ultraviolet radiation is emitted. An embodiment of a system for generating photons in accordance with the invention is shown schematically in FIG. 5. An acceleration structure 500 may correspond to the acceleration structure shown in FIGS. 2A and 2B, the acceleration structure shown in FIGS. 3 and 4, or in any other acceleration structure within the scope of the present invention. In the system of FIG. 5, acceleration structure 500 is a modification of acceleration structure 100 shown in FIGS. 1A and 1B and described above. Like elements in FIGS. 1A, 1B and 5 have the same reference numerals. Acceleration structure 500 includes concentric spherical electrode shells 112, 113 and 114, each of which is divided by an insulator 503 into electrode half shells. A pulse voltage source 540 is connected between inner electrode half shells 112a and 112b. A pulse voltage source 530 is connected between outer electrode half shells 114a and 114b and inner electrode half shells 112a and 112b. Acceleration structure 500 is enclosed within a housing 502 that defines a discharge chamber 504. A top aperture 140 of acceleration structure 500 is coupled through a screen 510 to a collection region 514 that is defined by an enclosure 516. Enclosure 516 contains collection optics 518 for relaying a photon beam 150 to a remote point of use. Screen 510 allows propagation of photons from discharge chamber 504 to collection region 514 but impedes flow of gas from discharge chamber 504 to collection region 514. A gas source 520 coupled to housing 502 supplies a working gas through inlets 522 and ports 134 in plenum 132 to acceleration structure 500. A bottom aperture 142 of acceleration structure 500 is coupled to a vacuum pump 524. An outlet 526 of vacuum pump 524 is connected to gas source 520 to form a gas recirculation system. The gas source 520 and the vacuum pump 524 are connected to housing 502 in a closed loop configuration that permits recirculation of the working gas through discharge chamber 504. Gas source 520 may include elements for removing impurities and particulates from the working gas. The system may include a detector 550 located in collection region 514, a control circuit 552 and a flow controller 554 for a feedback control of the rate of flow of the working gas into the discharge chamber 504 in response to a measured spectrum of the radiated photons. The system of FIG. 5 operates with first and second phases as described above in connection with FIGS. 2A and 2B. In particular, pulse source 530 applies a pulsed voltage between inner electrode half shells 112a and 112b and outer electrode half shells 114a and 114b, causing neutralized beams to be directed toward plasma discharge region 120. In the second stage, the plasma in discharge region 120 is heated and compressed by passage of a pulse of electric current. The neutralized beams form ionized tracks between cathode half shells 112a and 112b and plasma discharge region 120. Application of a pulse to cathode half shells 112a and 112b by pulse source 540 causes electrical current to flow along the ionized tracks through plasma discharge region 120, heating and compressing the central plasma. The plasma temperature and density rise to the point where the desired x-ray or extreme ultraviolet radiation is emitted. The radiation escapes the acceleration structure 500 as conical photon beam 150. A third embodiment of a photon source in accordance with the invention is shown in FIGS. 6A and 6B. Like elements in FIGS. 1A, 1B, 6A and 6B have the same reference numerals. The embodiment of FIGS. 6A and 6B differs from the structure of FIGS. 1A and 1B by the addition of an external electrode 600 for supplying an electrical heating current to plasma discharge region 120. A pulse voltage source 601 is connected between inner electrode shell 112 and outer electrode shell 114. A pulse voltage source 602 is connected between inner electrode shell 112 and external electrode 600. Insulators 603 and 604 electrically isolate the connections to electrode shells 112 and 114, respectively. External electrode 600 may have a cylindrical configuration and may be positioned in the bottom aperture of the acceleration structure in spaced relationship to plasma discharge region 120. During the first phase of operation, a pulse voltage V3 from voltage source 601 is applied between electrode shells 112 and 114. A discharge develops as described above, and the neutralized beams of the discharge pass through plasma discharge region 120 to form a plasma. At the same time, the passage of ions and energetic neutral atoms forms ionized tracks as described above. During the second phase of operation, a pulse voltage V4 from voltage source 602 is applied between external electrode 600 and electrode shell 112. The circuit is completed by conduction through the ionized tracks connecting electrode shell 112 and plasma discharge region 120 and through a glow region 605 between plasma discharge region 120 and external electrode 600. The current flows through the plasma in plasma discharge region 120, heating and compressing it. The plasma temperature and density rise to the point where the desired x-ray or extreme ultraviolet radiation is emitted. The plasma tends to be elongated in the direction of external electrode 600. The working gas extends to electrode 600 at approximately the same pressure as inside electrode shell 112. While there have been shown and described what are at present considered the preferred embodiments of the present invention, it will be obvious to those skilled in the art that various changes and modifications may be made therein without departing from the scope of the invention as defined by the appended claims.
	description	The United States of America may have certain rights to this invention under Management and Operating Contract No. DE-AC05-84ER 40150 from the Department of Energy. The present invention relates generally to radioisotope production, and, more particularly, to an apparatus and method for producing radioisotopes using linear accelerators. Recent technological advances, especially in the fields of diagnostic and therapeutic medicine, require ever-increasing quantities of radioactive isotopes. These radioisotopes are typically produced via irradiation of target materials in a small number of nuclear reactors or cyclotron accelerators across the globe. Isotopes produced in reactors are mainly from the neutron, gamma (n, γ) reaction (radiative capture). By contrast, cyclotrons bombard a target with a stream of heavy, charged particles (commonly protons). The highly desired radioisotopes are generally produced by a limited number of large facilities yielding a small variety of isotopes. Isotope selection is also limited, because the most useful isotopes often have short half-lives, making transportation a problem. The specific use of nuclear reactors also has the disadvantage of creating radioactive waste that is becoming increasingly problematic. A direct consequence of these limitations is that research and development in certain areas has stagnated because of the issues with production, transportation and economies of scale. Numerous industrial sectors would benefit from a compact, efficient, clean source of isotopes that is geographically close to the point of use, so as to take advantage of shorter half-life variants. It is therefore preferable to have a system and method that optimizes the production of the pertinent radioisotopes while minimizing the total energy needed for commercial production. The instant invention provides a solution to the foregoing problems. It is an object of the invention to provide a system and method that optimizes the production of certain radioisotopes while minimizing total energy requirements for such production. It is a further object of the invention to provide a system and method to produce a more diverse and reliable domestic supply of short-lived, high-value, high-demand isotopes at a cost lower than those extracted by reactors or positive-ion accelerators. It is also an object of the invention to provide a method of production of desired radioisotopes which does not result in the production of radioactive waste. The present disclosure provides a low cost, high output, eco-friendly system and method of radioisotope production based upon an energy recovery linear particle accelerator. An electron beam is accelerated in multiple 4.5 K SRF cavities and passed through a radiator where bremsstrahlung photons are created. The photons then hit a target for desired radioisotope production. The remaining electrons from the beam are sent back through the accelerator (SRF cavities) in deceleration mode, 180° out of phase, thereby regaining most of the RF energy prior to extraction at a beam dump. More specifically, one embodiment of the invention utilizes a 10 mA 50 MeV electron beam which is accelerated and energy recovered by 352 MHz SRF cavities at 4.5 K with 170 mm aperture. A cryomodule for the SRF accelerating cavities operates at 4.5 K providing, inter alia, substantially reduced technical complexity and power consumption. The present invention discloses a system and method for a radioisotope production facility based upon the premise that photonuclear reactions with bremsstrahlung photon beams from electron linacs can be used to generate isotopes of critical interest. Energy recovery linear accelerators (“ERL”) are increasingly the technology of choice for highly demanding applications. In energy recovery systems, more than 90% of the beam power is recycled, i.e., not deposited in a beam dump. The energy of the waste beam is therefore lower than the threshold for neutron production and the activation of shield components, thereby reducing both complexity and cost. ERLs offer superior beam characteristics well suited to isotope production. Of critical importance, as disclosed herein, energy recovery of an electron beam which has passed through a radiator can be accomplished. The Electron linacs proposed herein use a radiator to produce bremsstrahlung photon beams which interact with a target which creates isotopes of interest, as shown in FIG. 1. Bremsstrahlung or “braking radiation” is produced as charged particles (electrons) slowed down in their interaction with matter (the radiator). Referring now to FIG. 2, one potential embodiment of an ERL-based isotope production facility is illustrated schematically. The system begins with an injector, which provides the electron beam bunches, typically a laser-driven photocathode. This continuous wave electron beam is then accelerated by one or more SRF cavities. Typical accelerating gradients are 10-15 MV/m. The cavities are submerged in a helium bath within a cryomodule. The beam is steered and focused with magnets and beamline components until it is delivered to the target apparatus. The beam passes through a radiator composed of suitable material where bremsstrahlung photons are created. These photons in turn hit a target for radioisotope production. Instead of sending this beam to a dump (and wasting its energy and causing activation), the spent beam is steered back through the SRF cavities, 180° out of phase, so that it is decelerated and the energy recovered before being extracted to the beam dump. This recovered energy is returned to the RF structure, ready to accelerate the next bunch that passes through. For energy recovery to be efficient the cavities of a linac need to a) have a high accelerating gradient to give maximum acceleration per unit length of linac, and b) have low inherent losses in the accelerating structure. This naturally leads to the use of SRF over normal conducting copper structures because they possess both qualities. There is of course the added cost of the associated cryogenic systems that are required, but many studies suggest that during the operational lifetime of similar facilities the accumulated cost is significantly less. The instant ERL system is designed for operating at 4.5 K rather than the standard 2 K. This slightly elevated temperature means substantially reduced technical complexity and power consumption, and, moreover, can be achieved using commercially available refrigeration systems. The emphasis is on 4.5 K operation in order to utilize lower frequencies and optimize the design for low heat loss. The cost of a 400 W, 4.5 K helium refrigerator is approximately $4M. A 2 K plant would cost twice that, and be more complicated and costly to run. The surface resistivity of SRF materials scales with the square of the RF frequency and exponentially with temperature below the critical point. Therefore, operating in CW mode at 4.5 K requires lowering the frequency and reducing the gradients to achieve low heat loads and low operating costs. A useful result of using lower-frequency cavities is that the aperture size increases with decreasing frequency. The larger the aperture, the more disruption to the beam can be tolerated by the downstream components (including the cavities which harvest the energy of the ‘spent’ beam). In practical terms, a larger percentage radiator can be used, which produces more photons and hence reduces the time to produce a certain quantity of isotopes. The interaction between the electron beam energy, the beam current and the properties of the radiator material will define the photon spectra emitted via bremsstrahlung. As a consequence of this process, the features of the electron beam after the radiator will be degraded. The properties of photon spectra and the target material determine the yield of the reaction. A thin radiator composed of a suitable material, such as Tungsten, is used. An overly thick radiator may lead to excessive energy loss thereby impairing recirculation. Conversely, the radiator must be of sufficient dimensions so as to produce satisfactory amounts of bremsstrahlung radiation. Approximately 99% of the electron beam will pass straight through the radiator. A limit to energy recovery is set by the loss distribution through the radiator. FIG. 3 shows the experimental and theoretical loss distribution of a 53.61 MeV beam and a 74.63 MeV beam through a 1% and 6% Al radiator. It shows that the average energy loss in a 1% radiator is 0.378 MeV with a FWHM of 0.2 MeV. Studies at Jefferson Laboratory's free-electron laser indicate an acceptance limit on energy loss of 10-15% at 135 MeV. The spread in angles of a 50 MeV electron beam going through a 0.5% radiator is 26 mrad at 1 m. The radiator-induced increase in angular divergence must also be managed to permit efficient energy recovery, as discussed below. In order to demonstrate the potential performance of an electron linac based system, several important parameters must be looked at, e.g., the electron beam, the radiator performance (to understand the photon production rate) and the rate of isotope production in the target itself. For illustrative purposes, FIG. 4 sets forth various pertinent parameters in this system and the values for each with regard to one particular radioisotope, 67CU. It will be noted that while FIG. 4 addresses the production of 67Cu this system is not limited to the production of one specific isotope. Nonetheless, 67Cu is an example of an isotope produced through the photonuclear reaction 68Zn(γ, p)67Cu, and is ideally suited to be produced with an ERL. Using the foregoing parameters along with the standard linac operation parameters, the time it would take to produce 1 mCi of 67Cu is 142 seconds. Since the production rate is so high this system allows for the production of many curies of radioisotope before the decay rate becomes a problem. Using the foregoing rate of production, the estimated power cost would be $1.18 per mCi of 67Cu. These estimates demonstrate the potential cost savings of an ERL-based isotope production system. In the design of a high power ERL for isotope production, the electron beam will acquire large energy and angular spread through the interaction with the radiator target. The beam-target interaction will also generate a small fraction of secondary electrons which may act like a halo moving along with the core beam, which may cause cavity quenching and thus make high power operation unfeasible. To harness the high efficiency of ERL technology and use the ERL generated high power electron beam for isotope production, it is important to implement halo management schemes to control beam losses, and design the machine optics so that the exhausted beams with large energy or angular spread can be transported and dumped cleanly after energy recovery. With regard to the increase in the beam emittance and energy spread at the target, at E=50 MeV, for example, the electron beam going through an aluminum target with 0.5% radiation length can acquire 18 mrad rms angular spread, which is two orders of magnitude bigger than the typical beam divergence in a standard free-electron laser operation. Such large beam divergence poses significant challenges for the energy recovery in the linac and also for the beam transport to the dump. In general, the target arrangement, such as target type and thickness, determines the energy and angular spread of the spent electron beam at exit of the target. This spread, together with the beam spot size at target, further set specific requirements on the acceptance of the recirculation optics. Target heating is a potentially serious issue when using a high current electron beam in an ERL incident on a radiator where gamma photons are generated. With given electron beam power, the target heating tolerance (heat load per unit area) will set a lower limit for the beam spot size on the target. Meanwhile, the minimum achievable beam spot size on the target plays a defining role on the increase of beam emittance due to beam-target interaction, which will further determine whether the electron beam can be optically accepted by the downstream optics for the beam to be energy recovered in the linac. In many cases target heating can be circumvented by applying schemes, such as a cooling system or beam rastering systems, to increase the target heating tolerance. Basic considerations in the longitudinal and transverse match must be considered. For the longitudinal phase space, the interaction of the electron beam with the radiator will result in a large increase in beam energy spread while keeping the bunch length intact. The growth of longitudinal emittance can be minimized by having the beam reach minimum bunch length at the target position. This requires the longitudinal matching on the target similar to the matching at the wiggler for the FEL operation. For the transverse phase space, the beam divergence blows up dramatically from the beam-target interaction. To minimize the transverse emittance growth, the beam at the target needs to be matched to a minimal spot size in both the horizontal and vertical dimension. The beam after the target is mismatched to the downstream acceptance, and therefore the downstream acceptance needs to be reworked to rematch the output beam from target. The smaller the beam spot size at the target, the smaller is the emittance growth and the less difficult is the optics adjustment for divergence acceptance. This, however, is constrained by the minimum tolerable spot size at the target. For a given target arrangement, the lower limit of beam spot size is set by the target heat tolerance, while the upper limit is set by the maximum achievable optical acceptance for the beam line from the target to the back end of the linac after the beam is energy recovered. Taking all of the foregoing points into consideration, the design of a high power ERL based isotope production system requires attention to the impact of the beam spot size and target type and thickness on target heating, and on the maximum acceptance of the beam in downstream optics for energy recovery. With careful tuning of optics, the acceptance downstream of the target can be made such that energy recovery is feasible. Indeed, under ideal conditions, only around 10% of the electron beam energy is deposited in the beam dump and 90% or more of the electron beam energy is recovered. This method of isotope production, relying upon an electron accelerator, offers the potential for higher production yields for certain isotopes over more traditional methods. Further, since the linac energy is tunable, the resulting gamma energy can be optimized for isotope production cross sections. Highly focused (of the order of 100 μm) gamma rays can be generated, making it possible to produce radioisotopes with high activity in a small volume. Electron linacs, and in particular the system and method set forth herein, also offer other unique advantages over traditional techniques. One of the principal advantages is that such systems are reasonably simple devices to operate and maintain. This new and innovative approach using ERL technology has a number of advantages: i. Simplified separation of a desired isotope after production. ii. Energy recovery means greater overall (wall plug) efficiency. iii. Energy recovery allows a reduction in beam energies, to below the threshold of neutron production, lost to beam dumps. iv. Electron linacs are reasonably simple devices to operate and maintain. A production facility could be run by a relatively small group of technicians. v. Unlike methods relying upon a nuclear reactor, the ERL machine could be instantly powered down. vi. End of life decommissioning is simpler, cleaner and less costly than other technologies. vii. Smaller machine footprint allows flexibility in the placement and location of production facilities. The system and method disclosed herein provides efficiency and flexibility that can also be used to solve additional problems inherent in the current methods of isotope production. At this time, few options exist for getting isotopes to desired locations. Isotope selection is also limited, because the most useful isotopes often have short half-lives, making transportation a problem. A facility incorporating the apparatus and system set forth herein could be built in a hospital basement or in a distribution center within close geographic proximity of multiple hospitals and research establishments. The availability of new isotopes would open a frontier of innovative research and treatments and would help drive down production costs. While the invention has been described in reference to certain preferred embodiments, it will be readily apparent to one of ordinary skill in the art that certain modifications or variations may be made to the system without departing from the scope of the invention claimed below and described in the foregoing specification.
	claims	1. A boiling water nuclear reactor comprising:a core;a reactor pressure vessel containing the core;a reactor containment vessel including: a dry well containing the reactor pressure vessel, and a wet well containing a pressure suppression pool;a vent pipe connecting the dry well and the pressure suppression pool;a gravity-driven water injection pool arranged at a position higher than the core to hold boric acid aqueous solution therein;an emergency core water-injection piping system for causing the boric acid aqueous solution in the gravity-driven water injection pool to fall by gravity so as to be injected into the reactor pressure vessel in case of a reactor accident;a static containment vessel cooling system pool arranged outside the reactor containment vessel;a static containment vessel cooling system heat exchanger arranged in the static containment vessel cooling system pool;a dry well connection pipe connecting an upper part of the static containment vessel cooling system heat exchanger and the dry well;a gas vent pipe for discharging noncondensible gas in the static containment vessel cooling system heat exchanger into an inside of the pressure suppression pool;an intra dry well heat exchanger arranged in the dry well;an external heat exchanger arranged outside the reactor containment vessel;a dry well cooling water pipe connecting the intra dry well heat exchanger and the external heat exchanger in order to discharge heat obtained by the intra dry well heat exchanger at the external heat exchanger; anda boric acid aqueous solution heating device configured to heat the boric acid aqueous solution in the gravity-driven water injection pool by using part of the heat obtained by the intra dry well heat exchanger. 2. The reactor according to claim 1, wherein the gravity-driven water injection pool is arranged in the dry well. 3. The reactor according to claim 2, further comprising a drain pipe for discharging condensed water in the static containment vessel cooling system heat exchanger into the gravity-driven water injection pool. 4. The reactor according to claim 1, wherein:the gravity-driven water injection pool is arranged in the reactor containment vessel and partitioned from the dry well and the wet well; andthe nuclear reactor further comprises a first communication pipe holding a gas phase section in an upper part of the gravity-driven water injection pool and a gas phase section above the pressure suppression pool in the wet well in communication with each other. 5. The reactor according to claim 4, further comprising a drain pipe for discharging condensed water in the static containment vessel cooling system heat exchanger into the gravity-driven water injection pool. 6. The reactor according to claim 1, wherein:the gravity-driven water injection pool is held gas-tight and arranged above the reactor containment vessel; andthe nuclear reactor further comprises a second communication pipe holding a gas phase section above the gravity-driven water injection pool and the dry well in communication with each other. 7. The reactor according to claim 6, further comprising a drain pipe for discharging condensed water in the static containment vessel cooling system heat exchanger into a lower part of the dry well. 8. The reactor according to claim 1, wherein the gravity-driven water injection pool is held gas-tight and arranged above the reactor containment vessel; andthe nuclear reactor further comprises a third communication pipe holding a gas phase section in an upper part of the gravity-driven water injection pool and a gas phase section above the pressure suppression pool in the wet well in communication with each other. 9. The reactor according to claim 8, further comprising a drain pipe for discharging condensed water in the static containment vessel cooling system heat exchanger into a lower part of the dry well. 10. The reactor according to claim 1, further comprising an emergency dry well water-injection piping system for injecting the boric acid aqueous solution in the gravity-driven water injection pool into a lower part of the dry well in case of a reactor accident. 11. The reactor according to claim 1, wherein the pressure suppression pool can be used to fill a fuel exchange pool with water when replacing fuel loaded in the core. 12. The reactor according to claim 1, further comprising: a boric acid aqueous solution injection system including a boric acid aqueous solution injection tank arranged outside the reactor containment vessel to store pressurized boric acid aqueous solution of a quantity less than a quantity of water in the gravity-driven water injection pool and adapted to inject boric acid aqueous solution into the core from the boric acid aqueous solution injection tank. 13. The reactor according to claim 1, further comprising: a pressure-reducing valve arranged at the reactor pressure vessel at a position above the core so as to be capable of discharging steam from an inside of the reactor pressure vessel into the dry well. 14. An emergency core cooling system of a boiling water nuclear reactor, the emergency core cooling system comprising:a gravity-driven water injection pool to hold boric acid aqueous solution therein, the gravity-driven water injection pool being arranged at a position higher than the core;an emergency core water-injection piping system for causing the boric acid aqueous solution in the gravity-driven water injection pool to fall by gravity so as to be injected into a reactor pressure vessel containing the core, in case of a reactor accident;a static containment vessel cooling system pool arranged outside a reactor containment vessel that contains the reactor pressure vessel at a position above a pressure suppression pool;a static containment vessel cooling system heat exchanger arranged in the static containment vessel cooling system pool;a dry well connection pipe connecting an upper part of the static containment vessel cooling system heat exchanger and a dry well that contains the reactor pressure vessel and that is part of the reactor containment vessel;a gas vent pipe for discharging noncondensible gas in the static containment vessel cooling system heat exchanger into an inside of the pressure suppression pool;an intra dry well heat exchanger arranged in the dry well;an external heat exchanger arranged outside the reactor containment vessel;a dry well cooling water pipe connecting the intra dry well heat exchanger and the external heat exchanger in order to discharge heat obtained by the intra dry well heat exchanger at the external heat exchanger; anda boric acid aqueous solution heating device configured to heat the boric acid aqueous solution in the gravity-driven water injection pool by using part of the heat obtained by the intra dry well heat exchanger.
048805955	abstract	A process and an apparatus for cleaning nuclear reactor cooling water with cation exchange resin whose ion-exchanging groups have a bonding energy of not more than 300 KJ/mole are disclosed, whereby the radiation exposure of operators in an atomic power plant can be considerably reduced, and the waste ion exchange resin can be readily disposed.
	abstract	Disclosed is a system and method that tracks the usage of rented digital assets. A server includes an asset database that stores a digital asset, the title of the digital asset, and a server usage count for the digital asset. The computing device stores a rented digital asset and further stores an asset usage count list that includes the title of the rented digital asset and a usage count that indicates the amount of usage of the rented digital asset by the computing device. The server uploads the asset usage count list from the computing device and matches the title of the rented digital asset from the asset usage count list with the title of the same digital asset stored in the asset database. Further, the server adds the usage count of the rented digital asset to the server usage count for the digital asset in the asset database.
	summary
	abstract	An optical device for focusing synchrotron radiation light source is disclosed according to the present invention, so as to improve the uniformity of the light intensity of the emergent light, to increase the divergence the emergent light, and to restrain the synchrotron radiation higher harmonics. An outline generatrix of the optical device is a quadratic curve segment or a combination of a plurality of quadratic curve segments, the opening orientations of which are the same, and the optical device includes: a plurality of capillary bodies made of transparent material, wherein the capillary bodies in a center region have a solid construe; the capillary bodies in a periphery region located outside of the center region have a hollow structure.
051134233	description	DESCRIPTION OF THE PREFERRED EMBODIMENT FIGS. 1 and 2 illustrate the principle of the invention. In FIG. 1, wide, beam-shaped pulse 10 of partially coherent x-rays is shown being longitudinally sectioned into a group of narrower pulses, 20. The photons contained in pulse 10 have a certain degree of coherence transverse to the length of the pulse, but this coherence does not extend over the entire cross section of the pulse. In other words, along at least one direction, the transverse coherence length of the radiation is less than the width of pulse 10. After pulse 10 has been sectioned into a group of narrower pulses, 20, the transverse coherence length of each such pulse ore closely approaches the width of the pulses. Thus, the ratio of coherent radiation to total radiation is higher i.e. transverse coherence within each pulse has been improved. FIG. 2 is a schematic depicting pulses 20 after they have been delayed by varying amounts of time and re-ordered in serial fashion into a single pulsed beam, designated as 30, according to the method taught herein. The result of this re-ordering is that wide, less-coherent pulse 10 has been transformed into narrow, more coherent beam 30. Beam 30, due to its improved transverse coherence, is more useful in, for example, experiments involving the generation of interference patterns. Ideally, beam 30 will capture all of the energy originally present in pulse 10. FIGS. 3, 4 and 5 depict several schemes for sectioning pulse 10 into narrower pulses. The beam may be sectioned in either one or two dimensions. The number of sections required depends upon the application, with more sections being required for low-coherence input pulses. If required, the beam splitting process depicted in FIGS. 1-5 can be repeated in cascaded stages, with the split beams output from one stage being, in turn, split again in succeeding stages until the desired degree of width-narrowing is achieved. FIG. 6 is a schematic of a preferred embodiment of the invention. A pulse of input radiation is sent down beam pathway 600 to beam splitter 610. Split output pulses emerge from splitter 610 and pass via beam pathways 620 to a bank of static reflectors, designated collectively as 630, from whence the pulses are directed toward rotating mirror 640. The path lengths traveled from splitter 610 to mirror 640 are caused to be different for each beam pathway and reflector combination, such that only one pulse is arriving at mirror 640 at any given instant. The rotation speed of 640 and the angles of the beam tubes are adjusted such that each pulse arriving at the mirror is reflected into output light pathway 650. FIG. 7 is a perspective view of a beam-splitter suitable for use in this invention. Further construction details and theoretical background concerning this invention will now be presented. In order that radiation in a beam be sufficiently coherent, it has to satisfy certain conditions. Such coherence can be characterized for simple beam geometries by the coherence length, h.sub.z, of the beam and its two coherence diameters h.sub.x and h.sub.y, along the x and y axes respectively (the z axis is chosen to be parallel to the direction of beam propagation). If one requires that the coherence length, h.sub.z, be longer than a specified length, l.sub.z, then the beam must be sufficiently a) monochromatic, and b) collinear: a) Denote by .lambda. the wavelength of the radiation in the beam, by .DELTA..lambda. the full spread in .lambda., and by .lambda..sub.0 the average value of .lambda.. Sufficient monochromaticity requires ##EQU1## where f.sub.1 is a suitably chosen constant (to be specified below). b) Denote by .DELTA..theta..sub.i the full angular width of the beam along the i.sup.th axis (i=x,y). Collinearity will be sufficient, provided that ##EQU2## Here f.sub.2 is a suitably chosen constant (which will be specified later). c) Further conditions are imposed if one requires that the coherence diameters, h.sub.x and h.sub.y, of the beam along the x and y axes, be larger than some specified lengths l.sub.x and l.sub.y respectively. Let D.sub.si stand for the diameter of the beam source along the i=x,y axis (and assume that the beam axis is aligned with z). Consider coherence diameters at a distance L from the source, and assume that the beam is narrow, i.e. L>>D.sub.si, L>>h.sub.i ; i=x,y. Then the above requirement implies ##EQU3## Here f.sub.3 is a constant to be specified. PA1 C.sub..perp. can be increased only if at least one element of the beam optics is non-stationary. (S-1) PA1 the system of reflectors can not consist of only a single continuous mirror surface. (S-2) PA1 a) the photons are expensive to generate, PA1 b) the photon duty cycle is low, PA1 c) the photon intensity is one of the principal limiting factors in the experiment. As customary, we refer to the cross sectional area, A.sub..perp., corresponding to the two coherence diameters h.sub.x and h.sub.y, as the coherence area of the beam at a distance L from the source, and say that the coherence volume of the beam there is h.sub.z .multidot.A.sub..perp.. The number of particles in the coherence volume is the coherence number (sometimes also referred to as the degeneracy number). We introduce the following definitions: The "transversely coherent intensity", I.sub..perp., is the number of particles passing through A.sub..perp. per unit time. One can talk about instantaneous and average transversely coherent intensity. The transverse coherence number for a time interval .DELTA.t is the number of particles passing through A.sub..perp., during .DELTA.t, i.e. I.sub..perp. .multidot..DELTA.t. We denote the total particle beam intensity in the beam by I, and define the "degree of transverse coherence" or simply "transverse coherence", as EQU C.sub..perp. =I.sub..perp. /I. (4) Evidently, C.sub..perp. .ltoreq.1. The degree of transverse coherence is saturated when it reaches unity. At that point the transversely coherent intensity equals the total intensity, or, equivalently, the transverse coherence number for any .DELTA.t equals the total number of particles passing through a cross sectional area of the beam, oriented normally to the beam axis, during .DELTA.t. When conditions (1), (2) and (3) are satisfied, and when at the source all radiation is emitted in phase, then all radiation within a coherence volume h.sub.z .multidot.A.sub..perp. will be coherent, the exact degree of coherence depending on the constants f.sub.1, f.sub.2 and f.sub.3. In particular, when the distribution in wavelength, angle and point of emission are all uncorrelated, then difference between the phase of the radiation at any two points within the coherence volume will satisfy ##EQU4## ps The f is defined by the above equation. For sufficient coherence one usually requires EQU f.ltoreq.0.25. (5a) Conversely, if radiation from various points within the source is emitted with random phases, it follows from the symmetry of conditions (1)-(3) under the interchange l.sub.i .fwdarw. D.sub.si, (i=x, y), that one can perform interference experiments with such a source by allowing the beam to pass through two openings in a screen located in A.sub..perp., and observing the interference pattern behind the screen. Given any photon beam, it is always possible to increase the coherence volume: Using monochromators one can increase monochromaticity; while passing the beam through appropriate slits can improve collinearity and decrease the effective source diameters. By contrast, neither the transversely coherent intensity nor the (transverse) coherence number can be increased in this manner: Monochromators and slits operate by discarding photons with undesirable frequencies, angles, or points of origin. Slits would not be needed if the photon source itself had small enough emittance. For storage rings, that generally requires the reduction of the electron beam emittance, and there are definite engineering limits which can not be crossed at the present time. The method described here is capable of increasing the transverse coherence of the beam. In the limit it can saturate transverse beam coherence. DESCRIPTION OF THE METHOD a) First, the full beam is focused to have an angular divergence which does not contradict conditions (2). Assume that the cross section of the beam so obtained violates condition (3). PA0 b) Next, the beam is split into several component beams, altogether N.sub.c of them (FIG. 1). When the beam is an x-ray beam, this splitting can be accomplished by optical means. These beams all have cross sections consistent with inequality (3). If the full beam has radii .sigma..sub..gamma.i, (i=x,y), then the n.sup.th component beam has radii .sigma..sub..gamma.i.sup.(n) <.sigma..sub..gamma.i, n=1, . . . ,N.sub.c. (See FIG. 1). PA0 c) These component beams are allowed to travel along paths of different lengths to a common collection point P.sub.c, so that they arrive there in sequence, "stacked" one after the other. (See FIG. 2). PA0 From Eq (6) one then finds that the rotating mirror perimeter moves with a velocity ##EQU6## Here .epsilon..sub..gamma.y is the emittance of the photon beam along the direction m. On the other hand, the highest values v(r.sub.M) can reach are determined by the properties of the mirror material. For example, for uniform composition, denoting by .rho. and Y the density and tensile strength, respectively, ##EQU7## where F.sub.v is close to unity when h.sub.6 >>r.sub.m, referring to FIG. 12. Therefore, ##EQU8## For high grade steel one finds v(r.sub.M).ltoreq.6.10.sup.4 cm/s, so that when .epsilon..sub..gamma.y =10.sup.-9 rad m, and T.sub.p =10.sup.-6 s (similar to the values prevailing in synchrotrons), f.sub.c <6.10.sup.5. This limit is generally even more remote than the previous one. Finally, at point P.sub.c a rotating mirror directs all component beams through a port to the user. (For x-ray beams the mirror can be a simple reflecting surface, a multilayer, a crystal, etc.) The reconstituted beam emerging through the port will thus have not only the required angular divergence, but also the required cross section. If needed, adequate monochromatization (at any stage of the process) will then lead to appropriate coherence. As a result of this procedure, the beam will be transformed into a longer, but narrower one. When the beam cross section does not exceed the coherence area, then transverse coherence will be saturated, C.sub..perp. =1. Although the length of a pulse will increase, the longitudinal coherence length, h.sub.z, will not be increased by the method. On the other hand, there are techniques by which the method can be supplemented to increase h.sub.z. For example, longer undulators will cause an increase in h.sub.z, if the electron beam quality is good enough. A practical realization is shown in FIG. 6. In the case chosen here the decomposition pattern is one dimensional. There is no difference in principle between one and two dimensional decompositions, but the one dimensional case is easier to illustrate in a FIGURE such as this one. Furthermore, in many important cases one can reach complete coherence saturation by a one dimensional decomposition alone. One can prove in general, that by static means alone, one can never achieve an increase in the transverse coherence of an entire beam (as opposed to only a segment of it). Therefore, In FIG. 6 that element is a rotating mirror 640 designated by M.sub.r. After the beam multi-splitter at least one system of reflectors is needed to direct all beam components to P.sub.c. It can be shown that In FIG. 8 the system of reflectors, 630, consist of a sequence of disjoined mirrors, R.sub.i, i=1,2, . . . ,N.sub.c. Alternatively it may contain a grating structure or other equivalent discontinuous components, as illustrated in FIGS. 14, 15, and 16. Let us denote the length of a component x-ray beam by h.sub.i (i=1, 2, . . . , N.sub.c) so that it takes .DELTA.5.sub.i =(1/c) h.sub.j time for it to pass through any stationary optical element. To insure that each component beam will be clearly distinguished from every other one, it is necessary that the angular frequency of the rotating mirror by high enough: ##EQU5## Here l.sub.1i is the beam pathlength between R.sub.i and M.sub.r. The 2r.sub.M cos .phi. is the diameter of M.sub.4 parallel to the unit vector m. By definition, m is perpendicular to the projection of the axis of the beam reflected from M.sub.4, onto a plane perpendicular to the axis of rotation of M.sub.4 and is also perpendicular to the axis of rotation of M.sub.4. (See FIG. 13). The 2r.sub.Ri cos .sub.Ri is the diameter projected onto m, of R.sub.i when the beam reflected from M.sub.r hits R.sub.i ; the .sub.Ri is the angle of incidence of the beam on R.sub.i, and the components of the reflector system 630 are denoted by R.sub.i, i=1, . . . , N.sub.c. The .DELTA..theta..sub..gamma. is the full angular divergence of the reflected beam at M.sub.4, in the plane containing m. In Eq. (6) l.sub.i1 >>r.sub.M, r.sub.Ri is assumed. Let us denote by f.sub.c the factor by which the transverse coherence is increased as a result of coherence saturation. Then the time required to perform, e.g., a certain interference experiment will be reduced by this same factor. Large values of f.sub.c are desirable. In practice the maximum value of f.sub.c will be limited. One limitation on f.sub.c is related to the duty cycle, D, of the source. In designs such as the one shown in FIG. 6 one has to have EQU f.sub.c .ltoreq.1/D. (7) Since for high energy synchrotron sourced D.ltoreq.10.sup.-3, very significant f.sub.c values can be achieved before one has to deal with this constraint. Another limitation is imposed by the values of the angular velocity, d.phi./dt=.omega., that the rotating reflector element can achieve. To reach a certain f.sub.c value, the device must stack the i.sup.th component beam within the time .DELTA.t.sub.i. Assuming that all component beams have equal length, i.e., .DELTA.t.sub.1 =.DELTA.t.sub.2 =. . . =.DELTA.t.sub.nc =.DELTA.t, one finds .DELTA.t=T.sub.p /f.sub.c, where T.sub.p is the time which elapses between the onset of any two successive photon pulses generated by the source. Referring to FIG. 12, denote by r.sub.M the radius of the rotating mirror 1200 and by h.sub.zM its length. Let .sigma..sub..gamma.y and .sigma..sub..gamma.x be the radius of the photon beam along the direction m and perpendicular to it, respectively. Assume that .psi.=0, and the rotating mirror is large enough to intercept the entire photon beam incident on it: EQU r.sub.M .ltoreq..sigma..sub..gamma.y, (8a) EQU h.sub.ZM .ltoreq.2.sigma..sub..gamma.x /sin .alpha..sub.min.(8b) A third restriction on f.sub.c derives from the fact that the maximum difference in pathlength traveled by the various component beams, .DELTA..iota..sub.max, is related to the total pathlength across the instrument. For example, in the geometry illustrated in FIG. 10, one has EQU .DELTA.l.sub.max .apprxeq.1/2 l.alpha..sub.max.sup.2 ; l=l.sub.1 +l.sub.2,(12) where .alpha..sub.max is the maximum grazing angle of incidence. When one must have .alpha..sub.max <<1, one is restricted to .DELTA..iota..sub.max <<.iota.. To achieve any particular f.sub.c value, one needs EQU .DELTA.l.sub.max .gtoreq.f.sub.c T.sub.p Dc, (13) if the length of the individual component beams are assumed to be all equal, i.e., have the value cT.sub.p D. On the other hand, the beam optics must be so designed that the effective phase space occupied by the radiation is not significantly increased by random errors. In particular, the effect of random errors in angle, .delta..theta., due to mirror surface irregularities, should be small compared with the beam diameter. These effects have a value approximately equal to .iota..delta..theta., which requires EQU .iota..ltoreq..sigma./.delta..theta., (14) and limits .DELTA..iota..sub.max. This limitation can be significant. If so, it can be dealt with as described below. If .alpha..sub.max need not be <<1, this restriction is far less severe, and at the same time h.sub.zM as given in Eq. (8b) can be reduced. To evaluate the capabilities of the suggested approach, consider the SPEAR and PEP electron rings at Stanford. We assume that for SPEAR operating at circulating electron energy E.sub.e =1.5 GeV, the emittances are .epsilon..sub.x =1.125.times.10.sup.-7 rad m, and .epsilon..sub.y =1.125.times.10.sup.-9 rad m, and that in the region of photon generation the beta functions .beta..sub.x.sup.I =90 cm and .beta..sub.y.sup.I =8 cm; while for PEP operating at 4.5 GeV, .epsilon..sub.x =1.05.times.10.sup.-8 rad m, .epsilon..sub.6 =1.05.times.10.sup.-10 rad m, .beta..sub.x.sup.* =300 cm, and .beta..sub.y.sup.* =40 cm. From these the photon beam emittances, .epsilon..sub..gamma.x and .epsilon..sub..gamma.y, can be calculated at the source for both machines. The halflength of the electron bunches will be taken to be .sigma..sub.zo =5 cm for SPEAR, and 1.5 cm for PEP. These then are also the halflengths, .sigma..sub..gamma.zo, of the respective photon beam pulses generated. Table I lists the calculated values .epsilon..sub..gamma.x and .epsilon..sub..gamma.y for both machines for photons with energy E.sub..gamma. ; the coherence enhancement factor, f.sub.c ; the total length of the reconstructed resultant photon pulse L.sub.p ; the perimeter velocity of the rotating mirror v(r.sub.m)p as well as (1/2) .DELTA..theta..sub..gamma.y, r.sub.M, h.sub.z and .alpha.. The approximate length of the total optical path through the device can be estimated from .iota..gtoreq.2L.sub.p .alpha..sup.2 when .alpha.<<1, only the trivial condition .iota..gtoreq.L.sub.p remains. The procedure described previously and illustrated in FIG. 1 is well suited to explain the principle of coherence saturation. However, if .alpha.<<1, and the limitation of .iota. as discussed in connection with Eqs. 12 and 14 presents a problem, the design can be modified. In that case, rather than starting with a small (.DELTA./2) .theta..sub..gamma.y, it is preferred to first focus the beam with (.DELTA./2) .theta..sub..gamma.y sufficiently large compared to the random .delta..theta., so that the effect of the latter should become negligible. One pays for that either by having to deal with a significantly larger diameter beam later on, or by having to refocus the beam at least once before it reaches the rotating mirror. With a subsequent refocusing (1/2) .delta..theta..sub..gamma.y can eventually be reduced to its desired value. An alternative strategy which may be used in combination with the one just described, consists of decomposing the original beam in more than one step. In the first step each of the component beams are allowed to occupy a relatively large transverse phase space, large enough so that the relative increase caused by the random .delta..theta. is sufficiently small. In this step large .DELTA..iota..sub.1 can be induced, and in addition a certain .DELTA..iota..sub.2 space is left between successive component beams to allow the second step to take place. In the second step each component beam is considered to be the original beam, and further decomposed into subcomponent beams. In the second step it is sufficient to introduce .DELTA..iota..sub.2 difference in the optical path. When .DELTA..iota..sub.2 <<.DELTA..iota..sub.1. the benefits of this strategy become significant. One can also decompose the beam in more than two steps. In principle, the method proposed here can be used in conjunction with any noncoherent photon source. However, it should prove most immediately valuable when For high energy electron synchrotron radiation sources both a) and b) hold, and for interference experiments c) is also true. Therefore, coherence saturation should prove to be particularly valuable technique for such interference x-ray experiments. When the transverse coherence of a radiation beam pulse is increased by the method here described, the emittance of the beam along at least one transverse direction will be reduced in the process. That emittance reduction is often useful, even if it is not intended to use the beam in interference experiments. For example, it is sometimes desired to compress the length of a radiation pulse, in other words, it is desired to have substantially all of the radiation in the pulse arrive at a surface S during a time interval .delta.t which is shorter than the time duration, DT.sub.p, of the original pulse. Here S is assumed to be oriented substantially perpendicular to the axis of the beam pulse impinging on it. It is shown in the cited "Equitemporal X-Ray Optics" by Csonka 1986, that such length compression of a pulse can be achieved by dynamical optical means, i.e. when at least one optical element of the beam optics is non-stationary. Once the maximum speed of the moving optical element is given, the smallest .delta.t which can be accomplished by the method depends only on the emittance of the beam pulse whose length is to be compressed: .delta.t is smaller, if the emittance of the pulse is smaller along a chosen transverse direction. It is permissible to chose that transverse direction to be that along which the emittance is smallest. Therefore, smaller .delta.t can be achieved, if the emittance along at least one transverse direction is reduced. One can reduce the emittance by splitting the beam pulse into several components, and compressing the length of each component. That implies the complication of compressing several beam pulses. Alternatively, one may employ optical means to reduce the emittance along one transverse direction, while increasing the emittance along another transverse direction, which generally implies an increase of the beam diameter along the latter direction. That increase requires, in turn, larger moving optical elements, if the entire beam is to be compressed. At high speeds larger optical elements are more susceptible to instabilities. A third alternative, which may be used in conjunction with either or both of the above two, consists of decreasing the emittance along at lest one transverse direction while increasing the pulse length. Such is the case when transverse coherence is being increased by the method here described. Although the total pulselength is allowed to increase, the final achievable .delta.t will be smaller than for the original radiation pulse. The foregoing description of a preferred embodiment of the invention has been presented for purposes of illustration and description. It is not intended to be exhaustive or to limit the invention to the form disclosed, and, obviously, many modifications and variations are possible in light of the above teaching. It is intended that the scope of the invention be defined by the claims appended hereto.
039506517	summary	The present invention relates generally to equipment for defining the field, or target area, to be radiated in the medical treatment of patients and, more particularly, patients suffering from cancer, malign tumours, or similar diseases. As is well-known, when a patient is to be treated by radiation for the purposes above mentioned, the first step to be carried out by the medical experts is to define the desired target area. This is usually made by X-raying. The target areas vary from different patients not only in terms of location on the patient but also as far as their sizes and shapes are concerned. By way of example, in the treatment of lung cancer, it is not unusual to have a target area covering say one lung. Accordingly, such target areas often have very irregular contours. This circumstance coupled with the fact that it is necessary on the one hand to make sure that the radiation reaches all of the target area and, on the other, does not hurt adjacent tissues not affected by the disease, involves that the target area must be determined and defined with great accuracy. A third factual circumstance to be noticed in this context is that patients suffering from cancer have to be treated on repeated occasions, typically once a week for several months. The radiation is emitted by a radiation source mounted inside a therapeutic unit and it emerges therefrom through an aperture in a shutter carried by that unit. That aperture is generally of circular or rectangular cross-section and it is consequently necessary downstream thereof to modify the effective cross-section of the radiated beam so that it matches the contour of the target area. This is in prior art arrangements made by the use of blocks consisting of a radiation-absorbing material, generally lead. From what has been said above it is apparent that each time a certain patient is to be radiation-treated a number of such blocks have to be arranged so that the target area is properly defined. It is common practice to mark that area on the body of the patient and during the arrangement of the absorption blocks to have the apparatus emit ordinary light through said aperture so that the proper configuration of the blocks can be checked. A major disadvantage of the method above accounted for is that the block arrangement work has to be repeated on each treatment occasion. A second disadvantage is that the patient has to be present during the corresponding time period. This does not only mean that the patient has to be in the treatment room for a considerably longer period of time than that during which he is treated but also correspondingly reduces the availability of the most expensive equipment for the treatment of further patients. The main object of the present invention is to eliminate the disadvantages and shortcomings above referred to. A method for defining the target area to be radiated in the treatment of such a patient is, according to the present invention, characterised by the steps of mixing a heavy metal fine grain powder with a pressure-sensitive, non-curing adhesive so that a plastically deformable composition is formed, compressing a disc-shaped member consisting of said composition for the purpose of increasing its density and geometrical stability, providing said member with a through aperture the contour of which is congruent with that of said target area, and mounting said member between the shutter and the target area in such a position that the contour of said second aperture is located inside the cross-section of the beam. The invention does also concern a composition for use in the carrying out of the above defined method and a therapy unit constructed so as to operate according to the method.
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	abstract	In the present invention, a reactor power control apparatus of a natural circulation reactor comprises a power control apparatus for controlling reactor power and a pressure control apparatus for controlling reactor pressure. A power adjustment demand signal S4 is input from the power control apparatus into the pressure control apparatus. The pressure control apparatus controls degree of opening of a inlet port steam control valve provided in a moisture separation heater that is connected to a low-pressure turbine, based on the power adjustment demand signal S4.
061577014	claims	1. In an x-ray generating apparatus in which x-rays emitted from laser plasma formed by condensing and irradiating a laser beam on and to a target in a vacuum, are taken out from at least one side out of the laser beam irradiation side of the target and the side thereof opposite to said laser beam irradiation side, said x-ray generating apparatus comprising: a strong magnetic field generating means for generating a magnetic field component substantially vertical with respect to the target surface in the vicinity of said laser plasma, said magnetic field component being arranged to generate a magnetic force which acts directly on charged particles in said laser plasma to bend the tracks of said charged particles, causing said charged particles to be confined in a magnetic field formed by said magnetic field component, the magnetic flux of a strong magnetic field formed by said strong magnetic field generating means, is directed to the direction in which said laser plasma is generated, an x-ray supply object is disposed in a direction different from said laser plasma generating direction, and charged particles liable to be directed to said x-ray supply object are mainly confined in said strong magnetic field, wherein said target is disposed at the center of said strong magnetic field formed by said strong magnetic field generating means. said x-ray generating apparatus comprising: a strong magnetic field generating means for generating a magnetic field component substantially parallel with the target surface in the vicinity of said laser plasma, said magnetic field component being arranged to generate a magnetic force which acts directly on charged particles in said laser plasma to bend the tracks of said charged particles, causing said charged particles to be confined in a magnetic field formed by said magnetic field component, wherein the magnetic flux of a strong magnetic field formed by said strong magnetic field generating means, is directed to a direction different from the direction in which said laser plasma is generated, an x-ray supply object is disposed in said laser plasma generating direction, and charged particles liable to be directed to said x-ray supply object are mainly confined in said strong magnetic field. said x-ray generating apparatus comprising: a strong magnetic field generating means for generating a magnetic field component substantially vertical with respect to the target surface in the vicinity of said laser plasma, said magnetic field component being arranged to generate a magnetic force which acts directly on charged particles in said laser plasma to bend the tracks of said charged particles, causing said charged particles to be confined in a magnetic field formed by said magnetic field component, the magnetic flux of a strong magnetic field formed by said strong magnetic field generating means is inclined with respect to a surface of the target, an x-ray supply object is disposed in a direction different from said laser plasma generating direction, and charged particles liable to be directed to said x-ray supply object are mainly confined in said strong magnetic field. said x-ray generating apparatus comprising: a strong magnetic field generating means for applying a magnetic field to charged particles in said laser plasma, said strong magnetic field generating means being arranged to form a strong magnetic field for generating a magnetic force which bends the tracks of said charged particles in said laser plasma, causing said charged particles to be confined in said strong magnetic field, wherein said target is disposed at the center of said strong magnetic field formed by said strong magnetic field generating means. 2. In an x-ray generating apparatus in which x-rays emitted from laser plasma formed by condensing and irradiating a laser beam on and to a target in a vacuum, are taken out from at least one side out of the laser beam irradiation side of the target and the side thereof opposite to said laser beam irradiation side, 3. An x-ray generating apparatus according to claim 2, wherein the magnetic flux is directed in specific directions for effective confinement of charged particles. 4. An x-ray generating apparatus according to claim 2, further comprising a scattering particulate preventing means for intercepting low-speed scattering particulates. 5. An x-ray generating apparatus according to claim 4, wherein said scattering particulate preventing means comprises a high-speed mechanical shutter. 6. An x-ray generating apparatus according to claim 2, wherein the magnetic flux is directed in specific directions for effective confinement of charged particles, and wherein said apparatus further comprises a scattering particulate preventing means for intercepting low-speed scattering particulates. 7. An x-ray generating apparatus according to claim 2, wherein said target is disposed at the center of said strong magnetic field formed by said strong magnetic field generating means. 8. In an x-ray generating apparatus in which x-rays emitted from laser plasma formed by condensing and irradiating a laser beam on and to a target in a vacuum, are taken out from at least one side out of the laser beam irradiation side of the target and the side thereof opposite to said laser beam irradiation side, 9. An x-ray generating apparatus for generating x-rays by laser plasma formed by condensing and irradiating a laser beam on and to a target in a vacuum,
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053171411	description	DETAILED DESCRIPTION OF A PREFERRED EMBODIMENT FIGS. 1A and 1B are sectional views showing an apparatus according to the invention for use in semiconductor device fabrication, specifically in X-ray lithography for aligning an X-ray mask to a semiconductor wafer. An X-ray mask 30 is shown. X-ray mask 30 is conventionally a silicon nitride membrane. Means 32 for locating a feature on a surface of an object, specifically on a surface of a wafer, is mounted on and depends from a forward face of mask 30. A wafer 34 is shown mounted facing the forward face of mask 30. Wafer 34 is held in place on wafer chuck or mount 36. Mask 30 is provided with opaque patterned features, indicated at 38, as is conventional. In operation, X-rays are projected through mask 30 onto wafer 34. Patterning 38 blocks X-rays and thereby a pattern is projected onto wafer 34. With specific reference to FIG. 1A, means 32 for locating a feature on a wafer includes conductive pad 40, probe 42, and conductive line 44. Conductive pad 40 is provided on the forward face of mask 30 near an edge thereof. Conductive pad 40 is conventionally of chrome or tungsten, for example. Probe 42 is provided on a lower face of conductive pad 40. Probe 40 is a probe of a scanned probe microscope. A scanned probe microscope, as that term is used herein, includes all microscopes that operate by scanning a fine-tipped probe over the surface of an object and, by directly or indirectly monitoring interactions between the probe and the surface, to determine characteristics or properties of the surface. Such properties include, for example, topographic features and conductivity. Probe 42 of the embodiment of the invention depicted in FIG. 1A is a probe of a scanning tunneling microscope ("STM"). An STM operates by applying a voltage difference between the probe and the object surface sufficient to produce a tunneling current. Conductive line 44 is provided on the forward face of mask 30. Conductive line 44 is in electrical contact with conductive pad 40. Conductive line 44 extends to an edge of mask 30. A clip 46 is placed on mask 30 in contact with conductive line 44. Clip 46 is connected to appropriate conventional STM circuitry (not shown). Alternative electrical means may be provided for establishing a voltage differential between probe 42 and the wafer surface. Wafer 34 has provided thereon alignment mark 48. Alignment mark 48 as shown in FIG. 1 is a raised feature on the surface of the wafer 34. Alignment mark 48 may also be a valley or cut in the surface of wafer 34. In using an STM, it is essential to provide a conductive wafer surface, so that a voltage differential can be created between the probe 42 and the surface of wafer 48. A structure must be provided to achieve electrical contact between the surface of wafer 34 and the STM circuitry. For example, a conductive line 35, electrically connected to alignment mark 48 by an extension thereof on the surface of wafer 34, must be provided between the wafer and STM circuitry. A similar method can be employed on a dielectric surface, but an atomic force microscope, such as those described below, would be employed in place of an STM. Resists, such as resist layer 49, are generally insulators. Consequently, after forming of resist layer 49, a selected portion of resist layer 49 formed on and immediately adjacent to alignment mark 48, is removed by conventional techniques. Alignment mark 48 is desirably provided sufficiently distant from areas of the surface of wafer 34 where it is desired to form structures or devices that misalignment in removing the selected portion of resist layer 49, does not result in removal of resist 49 from areas where it is desired to form devices or structures. During the steps of forming of desired devices and structures on wafer 34, numerous layers will be formed on alignment mark 48. It is preferred that each such layer be removed prior to the next succeeding alignment step. It will be seen that the forward surface of mask 30 is sufficiently close to the surface of the wafer to permit the probe of the means for locating features on the wafer surface to be brought sufficiently close to the wafer surface to provide a tunneling current between the probe tip and the wafer surface. In X-ray lithography, it is conventional to hold the mask about 40 microns from the surface of the wafer. Thus, it is practical to provide a structure including a probe directly on the surface of the mask. It will be understood that, within the field of lithography, the structure illustrated in FIG. 1 may be used wherever the mask is provided in sufficiently close proximity to the surface of the wafer and is not limited to use on X-ray masks. According to FIG. 1B, there is shown a cross-sectional view of an apparatus 32' according to the invention for use in aligning an X-ray mask to a semiconductor wafer. The apparatus 32' of FIG. 1B is similar to the apparatus of FIG. 1A, but is adapted to provide alignment without removal of resist 49 from alignment mark 48. There is provided a substantially conventional probe 42' of an atomic force microscope ("AFM"). Support block 39 depends from mask 30. Conductive pad 41 depends from support block 39. Adjacent to conductive pad 41 and also depending from support block 39 there is provided bracket 43. Probe tip 45 depends from conductive pad 41. Arm 47 is provided on bracket 43. Arm 47 is provided immediately below the tip of probe tip 45, and is flexible, so that contact with the surface of wafer 34, or with the surface of resist 49, will cause arm 47 to deflect. Deflection of arm 47 changes a predetermined tunneling current between probe tip 45 and arm 47. It is particularly desirable that alignment mark 48 have topographical edges, or, in general, relief to the wafer surface, in the embodiment of FIG. 1B. In this embodiment, unlike the embodiment of FIG. 1A, resist layers, and other layers, formed on alignment mark 48 are not removed. In general, such layers will be highly conformal, even when numerous layers are formed. Thus, the edges of alignment mark 48 will be detected. It is not necessary, in the embodiment of FIG. 1B, that alignment mark 48 be electrically conductive. FIG. 2 is a block diagram showing interconnections in an apparatus according to FIG. 1A. The probe 42 and alignment mark 48 are both electrically connected to feedback control 64. Probe 42 and alignment mark 48 are in electrical contact to the extent that a tunneling current is maintained between them when a selected voltage differential is maintained by feedback control 64. Z-direction control 62 is electrically connected to an output of feedback control 64. The electrical connections in an apparatus according to FIG. 1B are similar. Probe tip 45 and arm 47 are both electrically connected to feedback control 64. Probe tip 45 and arm 47 are maintained in electrical contact to the extent that a tunneling current is provided between them when a selected voltage differential is maintained by feedback control 64. Feedback control 64 functions in a substantially conventional manner for feedback controls in STM devices. Thus, feedback control 64 monitors the tunneling current between the probe 42 and alignment mark 48. Feedback control 64 provides to Z-direction control 62 a control signal. This control signal causes the Z-direction control to effect relative movement of the probe and the wafer surface in the Z-direction. In the embodiment illustrated, the wafer is moved while the probe is stationary. An output of feedback control 64 is coupled to data processing apparatus 66. The feedback control provides Z-direction displacement data to data processing apparatus 66. The Z-direction displacement data is processed in accordance with conventional techniques to provide data regarding the location of alignment mark 48 on the wafer surface. The apparatus of FIGS. 1 and 2 is used as follows. Mask 30 is located relative to wafer 34 in accordance with conventional alignment techniques. The probe 42 is then used to determine the location of the feature or alignment mark 48. Then the mask and wafer are provided in preselected spatial relationship. In determining the location of alignment mark 48, the wafer is first raised, or moved in the Z-direction toward the mask. At the same time, the STM is activated. When a tunneling current is detected, control of the Z-direction apparatus is transferred to the feedback control. The probe is then rastered across a selected area of the surface of the wafer. A suitable algorithm may be provided for control of the relative movement of the probe and the wafer in the plane of the wafer surface during rastering, so that an appropriate area of the surface is scanned. The data is transferred to data processing apparatus 66, which creates a map of the scanned area. A suitable algorithm may be provided for identification of the location on the map of the scanned area of the surface of the alignment mark, and also a selected point on the wafer surface. The selected point for each mask must be that where, when the probe tip is at the selected point, registration is accurate. The mask will be in a preselected spatial relationship with the wafer when the probe tip is at the selected point. The location of the features on the mask relative to the location of the probe tip may vary among masks used on a single wafer as a result of slight differences in the process of forming the probe on the mask. Consequently, the relative locations of a reference feature on each mask and the probe tip must be previously ascertained. The relative locations are ascertained, for example, by performing metrology on the mask. For example, the completed mask may be scanned by a scanning electron microscope, using suitable calibration techniques. The data obtained from metrology is used to calculate a selected point for placement of the STM probe tip for each mask. The selected point for placement of the STM probe tip for each mask is defined relative to the alignment mark. The selected point for each mask may be stored in memory. After scanning the wafer surface, an image is generated. The selected point is located on the detected image. The present location of the probe on the detected image is also noted, and the selected location and the detected location are compared. A suitable algorithm may be developed for this comparison. The comparison step will provide a required displacement of the wafer with respect to the mask. The controller then operates the wafer chuck to move the wafer until the desired location is achieved. The movement of the wafer completes the positioning of the wafer and mask in preselected relationship. FIG. 3 illustrates in cross-section an apparatus for translational and rotational alignment of an X-ray mask with a wafer surface. The mask 130, wafer 134, and wafer mount 136, are as shown in FIG. 1 and described in the accompanying text. In this embodiment there is provided first means 132 for detecting the location of a feature on the wafer surface. This first detecting means 132 is identical to detecting means 32 of FIG. 1. Thus first detecting means 132 includes a conductive pad 140, a probe 142, a conductive line 144, and a clip 146. The wafer 134 has first alignment mark 148 provided thereon and conductive line 135 to provide an electrical connection thereto. Second means 150 for detecting the location of a feature on the wafer surface is mounted on and depending from the forward face of the mask 130. Second detecting means 150 is substantially the same as first detecting means 132. Second detecting means 150 is located near an edge of the mask 130. Second detecting means 150 is on an opposite side of mask 130 from first detecting means 132. Second detecting means 150 includes a second conductive pad 154 mounted on the lower surface of the mask. Second STM probe 156 is attached to a lower surface of conductive pad 154. Conductive line 158 is provided on the lower surface of the mask. One end of conductive line 158 is in electrical contact with conductive pad 154. The other end of the conductive line 158 is substantially at the edge of the forward face of the mask 130. Clip 160 clips to the edge of mask 130 in electrical contact with conductive line 158. Clip 160 is attached to standard STM circuitry. With reference to FIG. 4, there is depicted a block diagram of the circuitry of an apparatus in accordance with the embodiment of FIG. 3 for detecting the rotational alignment of a mask to a wafer. The circuitry relating to each of the two detecting means 132 and 150 is substantially identical to that explained in connection with the embodiment of FIGS. 1 and 2. First probe 142 and alignment mark 148 are electrically connected to first feedback control 162. First feedback control 162 is a substantially conventional feedback control for use in STM devices. An output of first feedback control 162 is connected to Z-direction control 170. An output of first feedback control is connected to data processing unit 166. Z-direction information is transmitted by first feedback control 162 to data processing 166. Second probe 156 is electrically connected to second feedback control 164. Alignment mark 152 is electrically connected to second feedback control 164. Second probe 156 and alignment mark 152 are electrically connected to the extent necessary to permit a tunneling current to flow between them. Second feedback control 164 is also a conventional feedback control for STM. The features in the surface can be observed by monitoring the tunneling current between the two probes when the water is rotated. The state must be manipulated until both probes provide the same relative mapping. It is within the capacity of one of ordinary skill in the art to design an algorithm to control the rotational movement of the wafer so as to obtain similar tunneling current responses from both probes. The apparatus of FIG. 3 is operated as follows. The mask and the wafer are roughly aligned according to conventional techniques. Then the location of first and second features on the wafer surface is determined, using the probes 142 and 156. In determining the location of the features, initially the first detecting means 132 and the second detecting means 150 are activated. The wafer is moved in the Z-direction to the mask until a proper tunneling current is established between the wafer surface and one of the two probes 142 and 156. The wafer is then rotated until a tunneling current is detected between the other of the two probes 142 and 156 and the wafer surface. The two probes are then rastered over a desired area of the wafer surface in accordance with a suitable algorithm. The data processing unit then generates maps of the two areas of the wafer surface. The two maps are then compared. A suitable algorithm may be developed for comparing the two maps. A comparison of the two maps will indicate whether the rotational alignment of the wafer and the mask is proper. If the two maps, after correcting for any misalignment of the probes in forming of the probes on the mask, show that the corresponding features on the two areas of the mask are in the correct relative positions, then there is no rotational misalignment. If corresponding features are not in correct relative positions in the two maps, then there is rotational misalignment. If rotational misalignment is detected, a suitable algorithm may be developed for calculating the degree of rotational misalignment. A suitable algorithm may also be developed for ascertaining the amount of rotation of the wafer required to correct the rotational misalignment. The next step is the positioning of the mask in preselected spatial relationship with respect to both alignment marks. In the step of positioning, either the wafer or the mask may in principle be moved. In the disclosed embodiment, the wafer is rotated. The wafer may be withdrawn in the Z-direction from the mask prior to rotation, in order to avoid damage resulting from contact between the probes and the wafer surface. It will be understood that this method and apparatus for detecting rotational misalignment is not limited to use with STM devices. Any type of scanned probe microscope may be used. For example, an AFM apparatus, as shown in FIG. 1B, may be employed, by modifying the apparatus of FIG. 1B to add a second AFM probe, and appropriate electrical connections. This method is also not limited to use in aligning masks with wafers. Any two objects may be aligned to high precision using this method and apparatus. A method for detection of rotational misalignment of a first object with respect to a second object may also be practiced using the apparatus of FIG. 1. According to this embodiment of the method, the alignment mark must have a shape such that its orientation can be ascertained. For example, a conventional cross-shaped alignment mark may be employed. The position and orientation of the alignment mark are determined and stored prior to practice of this embodiment of a method of the invention. The probe is rastered over the surface, including the alignment mark. The data processing apparatus generates an image of the alignment mark, which is compared to the stored position and orientation. Any differences between the stored orientation of the alignment mark and the orientation obtained by the probe indicate a rotational misalignment. A suitable algorithm may be developed for calculating the degree of rotational misalignment. Similarly, comparison between the stored position of the alignment mark and the position obtained by the probe will indicate any translational misalignment, as discussed above. A suitable algorithm may also be developed for rotating the wafer to achieve positioning of the mask in a preselected spatial relationship and orientation with respect to the alignment mark. This step of positioning results in a desired rotational alignment being achieved. FIG. 5 is a sectional view of an embodiment of the invention for use in semiconductor device fabrication, specifically for use in optical lithography. In the apparatus illustrated in FIG. 5, optical lithography is accomplished by projecting light through mask 502 to form a pattern. The light then passes through lens 504, which causes the light to become focused on the surface of wafer 506. A rigid support structure or frame 508 is provided to hold mask 502. Support structure 508 also supports first means 520 for locating features on the mask, second means 522 for locating features on the mask, first means 544 for locating features on the wafer, and second means 560 for locating features on the wafer. The structure and function of the means for locating features will be explained below. Mask 502 has alignment marks 512 and 514 located thereon. Similarly, wafer 506 has alignment marks 516 and 518 located thereon. The alignment marks may take any desired form. In the embodiment illustrated in FIG. 5, alignment marks 512 and 514 on the mask 502 are structures formed on the mask. The alignment marks may be formed simultaneously with the forming of a pattern on the mask. The alignment marks 516 and 518 on wafer 506 are illustrated as structures formed on the wafer surface. The alignment marks may be made of any material which can be conveniently used to form a structure on the wafer surface. The alignment marks on the wafer may alternatively be made, for example, as valleys cut into the wafer surface. Alignment mark 512 on mask 502, and alignment mark 516 on wafer 506 are precisely located with respect to one another so that, when wafer 506 is properly positioned with respect to mask 502, the displacement of arm 534 relative to alignment mark 512 is the same as the displacement of arm 554 relative to alignment mark 516. Similarly, when wafer 506 is correctly positioned with respect to mask 502, the displacement of arm 542 relative to alignment mark 514 is the same as the displacement of arm 570 relative to alignment mark 518 on wafer 506. The positioning of the alignment marks may be confirmed by conventional techniques, such as by performing metrology on the mask. The structure of the means for locating features will now be explained. First means 520 for locating features on the mask is a substantially conventional atomic force microscope. Piezoelectric block 524 depends from support structure 508. Conductive pad 528 depends from piezoelectric block 524. A probe 530 depends from conductive pad 528. Adjacent to conductive pad 528 there is provided a bracket 532, which also depends from piezoelectric block 524. A contact arm 534 is provided on bracket 532. Arm 534 is provided close to the tip of probe 530. As is conventional in atomic force microscopes, piezoelectric block 524 is operated in a conventional manner to move the structure with respect to support frame 508 and thus with respect to mask 502. Arm 534 contacts the surface of mask 502. Arm 534 and the tip of probe 530 are maintained, in accordance with conventional techniques in the art, a short distance apart, with an applied voltage between them, so as to cause a tunneling current to flow between the probe tip and the surface of arm 534. The tunneling current is detected in accordance with conventional techniques. As the arm is moved over the surface of the mask, changes in the mask topography cause the arm to move either closer to or further from the probe tip, thereby causing a change in the tunneling current. Alternatively, as is also conventional, a feedback circuit may be provided so that the piezoelectric block causes the apparatus to move vertically when the arm moves in a vertical direction, so as to provide a substantially constant tunneling current. Second means 522 for locating features on the mask is a substantially identical conventional atomic force microscope. Piezoelectric block 526, which depends from support frame 508, causes the arm and the probe to move with respect to the frame. A conductive pad 536 depends from piezoelectric block 526. A probe 538 depends from conductive pad 536. Immediately adjacent to conductive pad 536 a bracket 540 depends from the piezoelectric block 526. Arm 542 is mounted on the end of bracket 540, and extends directly below the tip of probe 538. Second means 522 for locating features on the mark operates in the same manner as first means 520 for locating features on the mask. It will be understood that either first means 520 for locating a feature on the mask, or second means 522 for locating a feature on the mask, may be used in accordance with the techniques described above, to ascertain the relative location of alignment marks 512 and 514 with respect to support structure 508. In addition, rotational misalignment may be detected by comparing the relative positions of marks 512 and 514. First means 544 for locating features on the wafer, and second means 560 for locating features on the wafer will now be briefly described. Both first means 544 for locating features on the wafer and second means 560 for locating features on the wafer are substantially conventional atomic force microscope devices, which are well known in the art. First means 544 for locating features on the wafer includes a piezoelectric block 546 depending from a lower arm of support frame 508. Piezoelectric block 546 has conductive pad 548 depending therefrom. Probe 550 depends from conductive pad 548. Bracket 552 also depends from piezoelectric block 546. Arm 554 is mounted on bracket 552 and extends so as to be directly below the tip of probe 550. Similarly, second means 560 for locating features on the wafer includes piezoelectric block 562 also depending from an arm of support structure 508. Conductive pad 564 is disposed depending from piezoelectric block 562. Bracket 568 is also disposed depending from piezoelectric block 562. Probe 566 depends from conductive pad 564. Arm 570 is attached at one end to bracket 568 and it extends so that it is immediately below the tip of probe 566. First means 544 for locating features on the wafer and second means 560 for locating features on the wafer are operated in a conventional manner of atomic force microscopes. It will be appreciated that both first means 520 for locating features on the mask and first means 544 for locating features on the wafer are attached to support frame 508 and are thereby maintained in a fixed spatial relationship with respect to one another. Second means 522 for locating features on the mask, and second means 560 for locating features on the wafer are also maintained in fixed spatial relationship with one another, as each is attached to support frame 508. When the wafer is brought into the proper position below first means 544 for locating features on the wafer, and second means 560 for locating features on the wafer, both first means 544 and second means 560 for locating features on the wafer are operated to determine the location of alignment marks 516 and 518 on the wafer with respect to support frame 508. The location of the alignment marks may then be compared, using suitable algorithmic techniques, to the detected location of the alignment marks 512 and 514 on mask 502. Any differences in alignment, resulting either from misalignment of the mask in its holder, or misalignment of the wafer, may be corrected by suitable movement of the wafer until the respective alignment marks are in registration to a desired degree of accuracy. A suitable algorithm may be provided to accomplish the necessary movement of the wafer. The wafer, for example, may be moved downward far enough to deactivate the means 544 and 560 for locating features on the wafer. The wafer may then be moved to correct for translational and rotational misalignment with respect to the mask. It will be appreciated, as explained above in connection with FIG. 1, that detection of rotational misalignment may be accomplished by the use of only one probe on the mask and one probe on the wafer. The wafer may then be moved in the upward direction to reactivate the first and second means 544 and 560 for locating features on the wafer. Then first and second means 544 and 560 for locating features on the wafer may be operated in accordance with conventional techniques to map the appropriate regions of the wafer surface to confirm that the alignment marks are located in registration with respect to the corresponding alignment marks on the mask. If the respective alignment marks are not in registration within the desired degree of accuracy, the process of adjusting the position of the wafer may then be repeated, and the resulting location tested by use of the first and second means 544 and 560 for locating features on the wafer, until the desired degree of accuracy is achieved. It will be seen that the embodiment of the invention illustrated in FIG. 5 is an advantageous technique for minimizing errors in registration in optical lithography. In successive exposure steps, the mask need not be located in precise registration with respect to its holder. Rather, the position of the wafer may be corrected so as to compensate for misalignment of the mask with respect to its holder. It will be understood that first and second means 544, 560 for locating features on the wafer and first and second mean, 520, 522, for locating features on the mask need not be in fixed spatial relationship, so long as the spatial relationship is known. FIG. 6 is a sectional view of an embodiment of the invention in which AFM devices are used to measure displacement between a mask holder and a mask and between the holder and a wafer surface. A mask holder 70 is shown. Mask 72 is supported along an edge by mask holder 70. A wafer 73 is shown disposed forward of mask 72 and holder 70. Means 74 for measuring displacement between the holder and the mask is provided depending from mask holder 70. Means 76 for measuring displacement between the holder and the wafer surface also depends from mask holder 70. Means 74 for measuring displacement between the holder and the mask includes an AFM device 81 movable on piezoelectric block 78. Piezoelectric block 78 is attached to and depends from mask holder 70. Piezoelectric block 78 has conventional electrical connections (not shown) to permit contraction and expansion when voltage differentials are applied in the x, y, and z directions. Support arm 80 depends from piezoelectric block 78. AFM device 81 is mounted on support arm 80. AFM device 81 has, as is conventional, a lever 84 for contacting a surface, and an STM probe 88. Support arm 80 is constructed and located so as to place the lever 84 of AFM device 81 in contact with a forward surface of mask 72. Probe 88 is located adjacent to lever 84 opposite from the mask surface. In operation, a tunneling current flows between lever 84 and probe 88. Bracket 82 is mounted on support arm 80. Lever 84 is attached at one end to bracket 82. Probe base 86 is also mounted on support arm 80 adjacent to bracket 82. Probe 88 is attached to probe base 86. Electrical connections (not shown) are provided to lever 84 and probe 88 in a conventional manner. Means 76 for measuring displacement between the holder and the wafer surface includes an AFM device 91. AFM device 91 is here shown movable on piezoelectric block 90. It will be evident that relative displacement between the AFM device and the wafer surface may be accomplished by movement of the wafer while the AFM device's position on the holder is fixed. Piezoelectric block 90 is attached to and depends from mask holder 70 adjacent to piezoelectric block 78 for the means 74 for measuring displacement between the holder and the mask. Piezoelectric block 90 has conventional electrical connections (not shown) to permit contraction and expansion when voltage differentials are applied in the x, y, and z directions. Support arm 92 is attached to and depends from piezoelectric block 90. AFM device 91 is mounted on support arm 92. AFM device 91 includes a lever 96 and a probe 100. Probe 100 depends from probe base 104, which depends from support arm 92. Lever 96 depends from bracket 102, which depends from support arm 92. Lever 96 is for contact with the wafer surface. Probe 100 is located adjacent to lever 96 opposite from the wafer surface. In operating a tunneling current flows between lever 96 and probe 100. Forces applied by the lever 96 on the wafer surface are also kept constant. The feedback control transmits z-direction displacement information to a data processing unit. This information is processed by a data processing unit in accordance with conventional techniques. The operation of the embodiment of the invention according to FIG. 6 will now be explained. The mask 72 is positioned in the mask holder 70 in accordance with conventional techniques. The relative displacement of the mask 72 and the mask holder 70 may be ascertained using means 74. The piezoelectric block 78 is operated in the z-direction to bring lever 84 in contact with the mask surface, so that a tunneling current is generated between the probe 88 and lever 84. The AFM unit 81 is then rastered on the surface of the mask. The piezoelectric block 78 is operated, in this embodiment, to raster the AFM unit 81. A data processing unit can then process the tunneling current information regarding the surface of the mask in the area that the AFM has rastered to produce an image or map of the surface. It is preferable in practicing this embodiment of the invention to have a stored reference image of the relevant sections of the mask surface. Such reference image may have been previously obtained by metrology on the mask. The stored reference image is then compared with the image obtained in scanning the surface. The features identified in the newly-obtained image can be compared to the reference image to determine the location of the AFM lever on the mask. It is only necessary to scan the mask surface until the scanned area can be identified on the reference image. A home position for the AFM is desirably designated on the reference image. A detected position for the AFM is then located relative to the reference home position. The relative displacement of the reference home position and the detected home position indicates the displacement of the mask from its proper position. The means 72 for detecting the position of the wafer relative to the holder is then activated. Initially, the lever 96 is brought into contact with the wafer surface. The z-direction orientation of the wafer and the holder is changed until a desired tunneling current is observed between the probe 100 and the lever 96. The AFM 91 is then rastered over the wafer surface. An alignment mark is located. When the AFM is at the alignment mark, and the mask is properly positioned in the holder, the mask and the wafer are properly aligned. Thus, if the mask is not properly positioned in the holder, it is necessary to effect relative displacement of the holder and wafer to obtain alignment. The wafer may be translated in the plane of the wafer surface to correct for the displacement of the mask from its proper position in the holder. As a result, the mask and the wafer are in alignment, to a very high accuracy. A process for forming a probe for a scanned probe microscope on a mask or other substrate will now be explained with reference to FIGS. 7 through 14. With reference to FIG. 7, there is shown in section a partial view of a surface of a substrate 10. The substrate 10 may be a glass plate of a mask for use in optical lithography. The substrate may be alternatively a silicon nitride membrane having a thickness of about 5000 angstroms, for use in X-ray lithography. Substrate 10 may also be a thick glass plate for use in optical lithography. A thin layer 12 of a metal is formed on substrate 10. The layer 12 will be referred to as the adhesion layer. The layer 12 is typically about 500 angstroms in thickness. The thin layer 12 is preferably of titanium tungsten. With reference to FIG. 8, there is depicted a sectional view similar to FIG. 10, but after forming of layer 14 on adhesion layer 12. Layer 14 is a tungsten layer. Adhesion layer 12 acts to seed the deposition of layer 14. Tungsten layer 14 is preferably about 500 angstroms thick. The desired pattern is then formed on the glass substrate 10 in the layers 12 and 14. In accordance with conventional techniques, a resist is applied on layer 14. A resist (not shown) is exposed, preferably by electron beam direct write techniques, and then developed. The layers 12 and 14 are then etched to provide a pattern on the glass substrate 10. The pattern provides a pattern for imaging on substrate 10 as well as locations for the addition of probe tips. Locations for bases for arms of AFM's (as explained below) may also be provided. Conductive lines to connect probe tip bases to an edge of the apparatus may also be provided. Layers 12 and 14 are etched using a CF.sub.4 -chlorine reactive ion etch (RIE etch) at 60 millitorr pressure at a gas flow rate ratio of CF.sub.4 to chlorine of 5 to 1, with a typical CF.sub.4 flow rate of 50 SCCM. At least one conductive base or pad 16 is provided as a result of the step of providing a pattern on the substrate 10. The pad 16 is shown in FIG. 9. A pad is provided wherever it is desired to provide a probe. A pad 16 comprises a tungsten surface layer 14. If it is desired to provide more than one probe, then more than one pad will be provided on a single mask. After formation of the pattern on the substrate 10, a layer of dielectric is formed on the substrate. The dielectric may be an silicon dioxide, and will generally be a glass. The dielectric layer 18 is shown in FIG. 10. As shown in FIG. 11, there is then formed on the dielectric layer 18 a spin-on-glass layer 20. The spin-on-glass layer 20 may be a siloxane resin. After the step of forming the spin-on-glass layer 20, a hole is etched through the spin-on-glass layer 20 and the dielectric layer 18 to the tungsten surface layer 14 of the pad 16. The result of this step of etching a hole is shown in FIG. 12. The hole 22 is formed in accordance with conventional techniques. A resist is formed on the layer 20. The resist is exposed and developed, leaving a hole in the resist where it is desired to form the hole 22. The layers 20 and 18 are etched in accordance with conventional techniques for etching spin-on-glass and glass dielectric layers. For example, CF.sub.4 -oxygen RIE etching techniques, at a CF.sub.4 to oxygen ratio of 5 to 1, with a CF.sub.4 flow rate of 75 SCCM at 60 millitorr pressure, may be used. The resist is then removed. The result of these steps is the structure shown in FIG. 12. After the step of forming the hole 22 through the spin-on-glass and dielectric layers through to the pad 16, the step of actually forming the probe on the pad is performed. Tungsten is selectively deposited by employing chemical vapor deposition techniques where WF.sub.6 is reduced with hydrogen at temperatures between 300.degree. C. and 450.degree. C. The hydrogen to WF.sub.6 ratio is 10 to 1 at 300 millitorr pressure, and a hydrogen flow rate of one liter per minute. The tungsten will only begin to grow using chemical vapor deposition techniques, on the tungsten surface layer 14. As a result of the conditions during which the chemical vapor deposition of tungsten takes place, the spin-on-glass layer 20 will begin to emit gases. These gases will be emitted sideways into the hole 22. The movement of these gases from the exterior to the center of the hole 22 will cause the tungsten to form preferentially in the center of the hole 22. As a result, the tungsten will form in the shape of a cone 24 as shown in FIG. 13. Thus the conical probe 24 is provided on the pad 16. It can be envisioned that any high-temperature polymeric material which could emit gases during the deposition process which suppresses CVD tungsten growth, could be used for this application. As shown in FIG. 14, after formation of the probe 24, the spin-on-glass layer 20 and the glass dielectric layer 18 are removed. These layers are removed in accordance with conventional stripping techniques using hydrofluoric acid, for example, with the underlying field nitride protecting the glass plate from attack. It should be noted that, with care, the spin-on-glass material can be chemically stripped while leaving some of the underlying oxide. After removal of the glass layers, the emitter or probe 24 may have a dirty surface. By a dirty surface, it is meant that various materials other than tungsten that were used in the initial tip formation steps have adhered to the surface of the emitter 24. It is thus desirable to create a clean surface by selectively depositing more tungsten on the probe 24 or cleaning the surface by hydrogen reduction in accordance with conventional techniques. Then the surface is heat treated in a hydrogen reduction atmosphere in accordance with conventional techniques. It is also known to those skilled in the art that probe tips can be chemically sharpened. With reference to FIGS. 15 to 23, there will now be explained a method for providing an AFM probe on a substrate. FIG. 15 is a top view of a patterned structure formed on substrate 10. FIG. 15A is a cross-section taken along line A--A of FIG. 15. Substrate 10 may be a conventional silicon substrate, or other dielectric. Formed on substrate 10 are conductive pads 12 and 25, conductive contacts 220 and 222, and conductive lines 13, 15. Conductive line 15 connects pad 25 and contact 220. Conductive line 13 connects pad 12 and contact 222. Contacts 220, 222, are preferably substantially larger in area than are pads 12, 25. Contacts 220, 222, may be, for example, 100 microns in length on each side, while pads 12, 25, may be, for example, 25 microns in length on each side. Conductive lines 13, 15 may be as narrow as desired. Contacts 220, 222, pads 12, 25 and lines 13, 15, are all preferably of a bilayer metallization of pure tungsten 14 on a titanium-tungsten alloy 16, and are formed by deposition, masking, and etching techniques as described above in connection with FIGS. 10-12. Referring to FIG. 16, which is a partial cross-section taken along line B--B of FIG. 15, there are depicted pads 12 and 25 after the step of forming of an STM probe 24 on pad 12. STM probe 24 is preferably formed in accordance with techniques described above. Dielectric layer 18 has been applied on pads 12, 25, as well as contacts 220, 222 (not shown) and lines 13, 15 (not shown). Spin-on-glass layer 20 has been applied on dielectric layer 18. A hole was opened through spin-on-glass layer 20 and dielectric layer 18 to expose a portion of the surface of pad 12, and probe 24 is formed by chemical vapor deposition of tungsten. Referring to FIG. 17, there has now been applied on spin-on-glass layer 20 and probe 24 an insulating layer 26. Insulating layer 26 may be a conventional dielectric. Referring to FIG. 18, a hole 27 is then provided by means of conventional patterning and etching techniques, through insulating layer 26, spin-on-glass layer 20, and dielectric layer 18, to the surface of second pad 25. FIG. 18A is a top view of the structures depicted in FIG. 18. Referring to FIG. 19, a metal 28 is then deposited over the insulating layer 26, and completely filling the hole 27. The second conductive pad 25 is then in electrical contact with the metal layer 28. The insulating layer 26 electrically isolates the probe 24 and the metal layer 28. Referring to FIG. 20, which is a partial, schematic, isometric view of a section of a structure formed according to a method of the invention, the metal layer 28 is now patterned and etched, according to conventional techniques, to provide a metal strip or arm 29 on the surface of the insulating layer 26. The metal arm 29 is in electrical contact at one end thereof with the second conductive pad 25 through the metal-filled hole 27. The opposite end of the metal arm 29 is disposed over the tip of probe 24. Referring to FIG. 21, which is a schematic, isometric view of a structure formed according to a method of the invention, the dielectric or insulating layer 26, the spin-on-glass layer 20, and the dielectric layer 18 are now stripped, by conventional techniques. The metal arm 29 remains supported at one end by a metal column formed by the metal that filled hole 27. At an opposite end, the metal arm 29 is disposed directly over the tip of probe 24. The metal arm 29 is disposed sufficiently close to the tip of probe 24 to permit the providing of a tunneling current between the arm 29 and the probe 24. Contacts 220, 222 are now exposed. It will be noted that a tungsten layer has been provided on contacts 220, 222, and connectors 13, 15. The tungsten layer is allowed to form on contacts 220, 222, and connectors 13, 15, for convenience in making the apparatus. However, it is not essential that a tungsten layer be provided on contacts 220, 222 and connectors 13, 15. The structure shown in FIG. 21 may be used as the probe of an AFM. Suitable electrical connections may be made to contacts 222 and 220. Connectors 13, 15 may be as long as desired, so that the steps of making electrical connections to contacts 222 and 220 will involve minimal risk of accidental damage to the AFM structure. There will now be described with reference to FIGS. 22 through 31, a further embodiment of a method of forming an AFM probe according to the invention. Referring to FIG. 24, there is shown in partial cross-section a substrate 300 with a release layer 302 disposed thereon. Substrate 300 may be, for example, a conventional silicon wafer surface. It will be understood from the following description of this method of making an AFM probe that the particular material of which substrate 300 is composed does not affect the process according to the invention. Release layer 302 may be, for example, a resist or other high-temperature organic material. Release layer 302 may be, by way of further example, a metal that is attacked, or dissolved, by a suitable etchant. In general, release layer 302 may be comprised of any material that can be removed, preferably by a solvent or etchant, with no or minimal damage to the metal layers that are applied in later steps, described below, according to the method. Referring now to FIG. 23, there is depicted a dielectric layer 304 disposed on release layer 302. Dielectric layer 304 may be any conventional insulating material. For example, dielectric layer 304 may comprise silicon dioxide. Dielectric layer 304 may be formed by any conventional technique, such as by CVD. In general, dielectric layer 304 may be comprised of any insulating material that may be suitably patterned, and that will be removed, such as by a solvent or an etchant, with little or minimal damage to metal layers that are applied in steps of the method described below. Referring now to FIGS. 24 and 24A, there is formed a hole 306 through dielectric layer 304. Hole 306 has sloping sides, so that hole 306 has the form of a truncated cone, with its base at the upper surface of dielectric layer 304. Any technique known to the art for forming hole 306 may be used. For example, dielectric layer 304 may be patterned by applying a suitable resist, exposing and developing the resist to expose a selected portion of dielectric layer 304 where hole 306 is to be formed, and then etching dielectric layer 304 by a technique that forms a hole having sloping sides. Referring now to FIG. 25, there has been deposited, on dielectric layer 304, and into hole 306, adhesion layer 305. Adhesion layer 305 may be formed of any conventional material which is used as an adhesion layer prior to the deposition of tungsten. For example, adhesion layer 305 may comprise aluminum or a titanium compound. Adhesion layer 306 may be formed by conventional techniques, such as by sputtering. Referring now to FIG. 26, there is deposited on adhesion layer 305, metal layer 308. Metal layer 308 is preferably formed of tungsten, by conventional chemical vapor deposition techniques, although physical vapor deposition can alternatively be used. It will be noted in FIG. 26, that tungsten layer 308 has completely filled the interior of hole 306 Adhesion layer 305 had already formed a thin layer on the interior surface of hole 306. Thus, it will be seen that metal layer 308 has a generally conical projection 309 formed thereon where hole 306 was located. Referring to FIG. 27, and FIG. 28, which is a cross-section taken along line B--B of FIG. 27, metal layer 308 is then patterned and etched, by conventional masking and etching techniques, to define a contact 320, and arm 322. Contact 320 may be, as shown, a substantially square structure having a large area. Contact 320 desirably has a large area, with a substantial length from its center to the nearest edge. Arm 322 is attached at one end to contact 320 and is integral with conical projection 309. Referring now to FIG. 29, which is a partial cross-section, similar to FIG. 28, there has been formed dielectric layer 310 on metal layer 308 and exposed portions of dielectric layer 304. Dielectric layer 310 is preferably formed of silicon dioxide. Dielectric layer 310 may be formed of any insulating material that will be removed, particularly by a solvent or etchant, with little or no damage to metal layer 308 and to a second metal layer applied further on in this detailed description. Dielectric layer 310 may be formed by conventional techniques, such as chemical vapor deposition techniques. Referring now to FIG. 30, there has been deposited on dielectric layer 310 second metal layer 312. Second metal layer 312 may be any conventional metal, such as tungsten, with an adhesion layer of, for example, titanium tungsten. Second metal layer 312 may be formed by any conventional metallization technique, such as by sputtering. Subsequent to deposition, second metal layer 312 is patterned. After the patterning step, second metal layer 312 has the form shown in FIGS. 30A and 30B. FIG. 30A is a top view of the structure formed according to this embodiment of the method of the invention after the step of patterning. FIG. 30B is a cross-sectional view taken along line B--B of FIG. 30A. It will be seen that a metal arm 313 is formed, which is disposed directly over arm 309 of metal layer 308. A large contact 315, from which arm 313 extends, is also defined. Contact 315 is preferably disposed directly over contact 320 and may have substantially the same shape and dimensions as contact 315. Referring now to FIG. 31, which is a partial cross-section similar to FIG. 30B, there has been attached to patterned second metal layer 312 second substrate 314. Second substrate 314 may be made of any suitable material. For example, second substrate 314 could be a second silicon wafer. By way of further example, second substrate 314 could have a generally ring-like form. Substrate 314 is attached to metal layer 312 by techniques known in the art. For example, a standard adhesive could be used, or heat contact techniques known in the art could be employed. Referring now to FIG. 31A, release layer 302 has been removed by an appropriate solvent or etchant that dissolves the material comprising release layer 302, but does not substantially attack the metals of metal layers 308 and 312. As substrate 300 is no longer attached to either metal layer, substrate 300 is removed. Referring now to FIG. 32, which is a partially schematic isometric view of an AFM probe structure 316 fabricated in accordance with the method of the invention, and to FIG. 32A, which is a cross-section taken along line A--A of FIG. 32, a portion of dielectric layer 310 has been removed, preferably by use of a suitable solvent which attacks silicon dioxide but does not substantially attack the metals of first metal layer 308 and second metal layer 312. The time that the solvent is permitted to be in contact with the structure is controlled. The time is selected to be sufficiently long that dielectric 310 is entirely removed from arms 313 and 320. The time is also sufficiently short that a substantial width of dielectric 310 remains intermediate contacts 315 and 322. Dielectric layer 310 remains between contacts 315 and 322. AFM probe structure 316 includes metal arm 313, from which probe tip 309 depends. Probe tip 309 may be sharpened, by conventional chemical techniques known in the art of forming AFM and STM probes, after the step of dissolving release layer 302. Metal arm 313 depends from contact 315. Intermediate metal arm 313 and metal arm 320, there is defined a gap. Thus, dielectric layer 310 serves to support metal arm 313 a preselected distance from metal arm 320, and maintains contact 315 and contact 322 electrically isolated from one another. In operation, appropriate leads are provided to attach contact 315 and contact 322 to a processing apparatus for an AFM (not shown). A desired capacitance is obtained between arm 313 and arm 320, by establishing the relative distance between them. Tip 309 is then brought into contact with a surface, and scanned across the surface in accordance with conventional techniques. Forces between the surface and the tip of probe 309 caused by topographical variations on the surface on the order of Angstroms force arm 313 to be deflected upward, thereby changing the distance between arm 313 and arm 320, which causes a corresponding capacitance change. The capacitance change is detected by appropriate measuring instruments. It will be appreciated by those of skill in the art that various dimensions may be selected for an apparatus made according to the above method. By way of example only, the thickness of first metal layer may be approximately 2 microns, the thickness of dielectric layer 310 may be approximately 500 Angstroms, and the thickness of metal layer 312 may be approximately 2 microns. There will now be described, with reference to FIGS. 33 through 41, an alternative method according to the invention of making an AFM probe assembly. Referring now to FIG. 33, there is shown substrate 400 having release layer 402 formed thereon. Substrate 400 may be, for example, a conventional single-crystal silicon substrate. As will become apparent from review of the detailed description of this embodiment of the method of the invention, the composition of substrate 400 is not important to the method of the invention. Release layer 402 may be, for example, a resist or a high-temperature polyamite. Release layer 402 may also be, by way of further example, a metal that is attacked by an etchant or solvent that does not attack the metal of which the additional metal layers described below are formed. In general, release layer 402 may be composed of any material that can be removed, especially by a solvent or etchant, with no damage, or minimal damage, to metal layers made subsequently. Referring now to FIGS. 34 and 35, there is shown metal member 404 formed on release layer 402. Metal member 404 is preferably formed by conventional metallization, patterning and etching techniques. For example, a metal layer may be deposited by chemical vapor deposition on release layer 402. A conventional resist may be applied on the deposited metal layer, which resist may be exposed and developed through conventional techniques. The resist may then be used as an etch mask in etching the metal layer to form metal member 404 with support member 405, which is substantially larger than metal member 404. Metal member 404 has a preselected, generally elongated rectangular shape with a large pad 405 attached at one end thereof. The dimensions of metal member 404 may be, by way of example, 10 micrometers wide and 50 micrometers in length. Referring now to FIG. 36, there is shown a dielectric layer 408 formed over metal member 404, and the exposed portions of release layer 402. Dielectric layer 408 may be provided by conventional processes. In a preferred embodiment, dielectric layer 408 is composed of silicon dioxide, and is deposited by CVD techniques. In general, dielectric layer 408 may be composed of any insulating material that can be removed by a solvent or etchant that will not substantially attack metal member 404. Referring now to FIG. 37 and FIG. 37A, there is shown a roughly conical depression 410 formed in dielectric layer 408. Depression 410 is disposed over metal member 404 at a selected position. Depression 410 does not extend downward in dielectric layer 408 to the surface of metal member 404. Thus, a thin portion of dielectric layer 408 separates the surface of depression 410 from metal member 404. Depression 410 may be formed by conventional masking and etching techniques that are well known to one of ordinary skill in the art. For example, an erodable mask may be used. Referring now to FIG. 38, there is shown second metal layer 412 formed over dielectric layer 408. Second metal layer 412 is preferably of tungsten. Second metal layer 412 may be formed by conventional techniques. For example, a thin adhesion layer of aluminum or a titanium alloy may be formed on dielectric layer 408 by conventional techniques, such as by sputtering. Tungsten may then be deposited on the adhesion layer by conventional techniques, such as by CVD techniques. Second metal layer 412 fills depression 410 to define a metal projection 409. Referring now to FIG. 39, there is shown second substrate 414 disposed on an upper surface of second metal layer 412. Second substrate 414 may be any desired material, and of any desired shape and size. For example, second substrate 414 may be a conventional single-crystal silicon wafer. Second substrate 414 may be attached to metal layer 412 either by conventional adhesives or by conventional heat contact techniques. Such techniques are well known to those of skill in the art, and will not be explained here. Referring to FIG. 40, there is shown in cross-section a structure after a step of removal of release layer 402. Release layer 402 is removed by use of an appropriate solvent or an isotropic oxygen plasma which does not tend substantially to attack the metal layers. As the metal layers are no longer attached to substrate 400, substrate 400 is no longer shown. Referring now to FIG. 41, dielectric layer 408 has been partially removed. Dielectric layer 408 is partially removed by a suitable etchant or solvent. It is within the skill in the art to select an appropriate solvent or etchant that will attack the materials of which dielectric layer 408 is made, without attacking, or with only minimal attacking of, metal member 404 and pad 405, and metal layer 412. Dielectric layer 408 remains intermediate pad 405 and metal layer 412. It will thus be seen that a structure is provided, which is attached to substrate 414, and comprises metal member 404 and metal layer 412, both attached to the remaining portion of dielectric layer 408, and disposed a selected distance apart. Depression 410 has been filled with metal in the forming of metal layer 412, and thus is a tip 409. Referring now to FIG. 42, there is shown the structure of FIG. 41, after the sharpening of tip 409 to provide tip 409 with a conical form having a very fine apex. Sharpening of tip 409 may be accomplished by chemical techniques that are well-known in the art. It will be seen that the structure shown in FIG. 42 may be employed as an AFM probe. Metal member 404 can be brought into contact with a surface to be scanned, and can be translated across the surface. As a result of repulsive electrostatic forces between the lower surface of metal member 404 and the object or target surface, metal member 404 will be caused to flex upward toward tip 409. By attachment to suitable AFM circuitry, a slight voltage differential may be established between metal layer 412 and metal arm 404. A tunneling current may be established between tip 416 and metal arm 404 by proper setting of the voltage differential, according to techniques well known in the art of constructing and operating an AFM. Upon deflection of metal arm 404, the tunneling current will change. Thus, the structure shown in FIG. 42 may be used as an AFM probe. Other techniques that can be used to measure deflection may also be employed. It will be appreciated by those of skill in the art that various dimensions may be selected for an apparatus made according to the just-described method. By way of example only, the thickness of dielectric layer 408 may be about 500 Angstroms. The thickness of metal arm 404 may be about 2 microns. It will be appreciated that there are considerable variations that can be accomplished in a method and apparatus of the invention without departing from its scope. As a result, although preferred embodiments of a method and apparatus of the invention have been described above, it is emphasized that the invention is not limited to a preferred embodiment and there exist other alternative embodiments that are fully encompassed within the invention's scope, which is intended to be limited only by the scope of the appended claims.
061577030	abstract	An x-ray beam hardening filter is disclosed. The x-ray beam hardening filter comprises a support member and a beam hardening sheet, the beam hardening sheet having a multidimensional array of regularly spaced apertures. The apertures are configured to have an x-ray transmissive quality. An actuator, engaging the support member, is capable of moving the multidimensional array of apertures into or out of a path of an x-ray beam, thereby selectively introducing varying levels of x-ray energy filtration. In one embodiment, multiple layers of beam hardening sheets are added to the x-ray beam hardening filter to create additional levels of x-ray energy filtration. Advantages of the x-ray beam hardening filter include the relatively small distance the x-ray beam hardening filter must move in order to absorb the incident x-ray beam, the ability to introduce varying levels of x-ray filtration, and the compact structure of the x-ray beam hardening filter.
043199607	summary	BACKGROUND The present invention relates to nuclear fuel and in particular to a rack for the storage of nuclear fuel assemblies. Fuel for commercial nuclear fission reactors typically includes a stack of enriched uranium dioxide pellets contained within a tube called a fuel rod. A fuel assembly includes about 200 to 300 fuel rods held in a square array by spacer grids and upper and lower end fittings. It is common for nuclear fuel assemblies, when not in use in the reactor steam cycle, to be stored in pools of water at the nuclear reactor site or at another site. Fuel assemblies are stored both before and after they are used to produce energy in the reactor. Typically, nuclear fuel assemblies are stored under water at reactor sites in some sort of rack structure, spaced from each other a sufficient distance to preclude a sustained nuclear fission chain reaction. One type of rack uses a pair of horizontal frames supporting the fuel assembly at the top and bottom in a vertical attitude. In another the fuel assemblies are inserted into cans. The top and bottom ends of the cans are supported by some form of horizontal frame. These types of structures have significant flexibility, a disadvantage in that it permits them to respond to seismic or other excitations raising the possibility of damage to the stored fuel assemblies and potential leakage of radioactive materials therefrom. The lack of nuclear fuel reprocessing facilities has forced many utilities and fuel assembly manufacturers to increase the number of fuel assemblies stored in their storage pools. The most common way to increase capacity of a storage pool is to use a more closely spaced fuel assembly array. A sustained nuclear fission chain reaction is precluded in the closely spaced array by positioning a neutron absorbing material between the stored fuel assemblies. Current closely spaced fuel assembly rack designs require expensive fabrication techniques and inefficient use of structural material. The present invention is a nuclear fuel assembly storage rack concept consisting of four identical plates joined in a novel arrangement to form a nuclear fuel assembly storage cell. A number of these storage cells can be joined in an array to form a modular rack structure of variable array size depending on the individual plant application and needs. The storage cell is made of stainless steel plates which can be assembled on a mandrel to achieve required dimensional control and square corners on the inside of the cell. The plates can be joined together by either riveting or welding the easily accessible flange areas. Since the plates are identical, the advantages of part standardization are present. Material procurement, tooling, and fabrication are all simplified. The shape of the plates and the arrangement of cells create a cavity between adjacent cell walls into which the neutron absorbing material can easily be placed. The cells can be joined together to form a stiff array to permit the natural frequency of the structure to be made sufficiently high thus avoiding large seismic response. Seismic loads that are generated can be carried from cell to cell, rather than through an external framework, reducing the amount of structural steel needed. The rack structure can be designed to be free standing; ie: no braces between rack and pool walls thus eliminating structural steel bracing. It is an object of the invention to provide a nuclear fuel assembly storage rack. A further object of the present invention is a storage rack utilizing a single standardized structural part resulting in lower unit fabrication and reduced tooling costs. Another object of the invention is a storage rack yielding the foregoing advantages and which is simple to fabricate, thereby reducing welds or riveting, simplifying dimensional controls and reducing inspections. Another object of the invention is a storage rack yielding the foregoing advantages and which is modular in construction and thus may be easily adjusted to any storage pool configuration or plant arrangement. Other objects and advantages of the present invention will be readily apparent from the following description and drawings which illustrate the preferred embodiment of the invention. SUMMARY OF THE INVENTION The present invention involves a nuclear fuel assembly storage rack including a plurality of identical plates affixed together to form fuel assembly storage cells. The cells are arranged in an array and interengaged to form the storage rack.
	abstract	To use 99mTc as a raw material for a radioactive medicine, a very small amount of 99mTc in the high concentration Mo(99Mo) solution is purified and recovered with high yield without contamination of 99Mo. 99mTc with high purity is recovered by forming a high concentration Mo(99Mo) solution which contains radionuclides 99Mo which is the parent nuclide of 99mTc used for the radioactive medicine and the raw material for its labeled compound, forming a high concentration Mo(99Mo) solution which contains radionuclides 99Mo and 99mTc by generating 99mTc to a radioactive-equilibrium state, getting 99mTc in the high concentration Mo(99Mo) solution adsorbed to activated carbon selectively by feeding the solution to an adsorption column which has activated carbon, and undergoing desorption and purification treatment of 99mTc with a desorbent from the activated carbon to which 99mTc is adsorbed.
050948008	summary	BACKGROUND OF THE INVENTION The present invention relates to a press for compacting or compressing elongated radioactive structural elements, such as fuel cell skeletons, including a horizontally disposed press line that comprises a press shaft provided with a cover, a press ram operated via a hydraulic cylinder, and a removable counterpunch in place of which, after a compressing process has been completed, is inserted a transfer shaft into which is introduced the highly compressed pressed object via further advancement of the press ram. The present invention also relates to a method for operating such a press. A press of this general type was disclosed in German Offenlegungsschrift 38 02 966. The press ram thereof is driven by a long hydraulic cylinder. Via the press ram, a fuel cell or element skeleton is compressed in the longitudinal direction against the base of the press shaft. After the compression, the pressed object is pushed into a transfer shaft with which it is conveyed out of the press. The transfer shaft extends about the pressed object, so that broken pieces of skeleton remain secured, and an expansion of the volume in the radial direction due to expansion of the pressed object is prevented. The ability to keep the pressed object within certain dimensions is advantageous for further operating stages. Due to the necessary long hydraulic cylinder, a drawback of this proven press is its great structural length, which also makes a remote handling more difficult. It was furthermore discovered that toward the end of the compressing process, smaller particles can escape from the press line out of the gaps that necessarily result between the end face of the press body and the end face of the counterpunch during assembly. Due to the resilience of the pressed object after conclusion of the compression process, the pressed object can swell in the axial direction and thereby prevent removal of the counterpunch. German Offenlegungsschrift 34 44 570 proposes an apparatus with which elongated radioactive parts that have the same geometry are cut into pieces and are subsequently introduced into a receptacle. The length of these pieces corresponds approximately to the height of the receptacle. By means of a packing mechanism, pieces that are already in the receptacle are pressed to the side in order to provide space for further pieces. This mechanism is designed for pieces having a very specific geometry; it is not suitable for compacting or compressing elongated radioactive parts having varying geometries, such as non cut-up fuel cell skeletons. German Offenlegungsschrift 33 00 359 furthermore proposes an ejection mechanism for compactable material, where the material, which is in pieces or is granular, is compressed into plugs in a compression chamber. In so doing, the degree of compression can be set by changing the counter holding force of the closure means for the compression chamber. The use of this ejection mechanism is limited to loose material that can be poured. It is therefore an object of the present invention to provide a press of the aforementioned general type that has a shorter overall length, that to a large part can be handled or controlled remotely, and that during the compression process as well as during introduction of the pressed object into the transfer shaft, prevents any particles from escaping out of the press line, with this press nonetheless assuring that the unavoidable swelling of the pressed object in the axial direction will have no impact upon the removal of the counterpunch or the insertion of the transfer shaft.
047708433	claims	1. A method of controlling the stability of the fuel assemblies in a boiling water reactor core comprising the steps of: (a) monitoring on an on-line basis the values of selected reactor parameters incuding, flow, inlet subcooling, control rod position, and pressure; (b) determining on a real time basis from the selected reactor parameters, the power generated in, and the axial power distribution for, each fuel assembly; (c) selecting fuel assemblies with axial power peaking occurring below a predetermined location as determined from said axial power distributions, and with a power level above a selected value; (d) calculating on-line in a digital computer a stability index only for said selected fuel assemblies; and (e) generating a human readable representation of the stability indexes in real time for the selected fuel assemblies. iteratively calculating the stability index for said least stable fuel assembly incremental changes in a selected one of the axial power distribution and flow parameters until said stability index for the least stable fuel assembly is reduced at least to said preselected stability limit; and operating reactor controls to implement the cumulative change in the selected one parameter to bring the stability index for the least stable fuel assembly down to at least said preselected limit. (a) measuring on an on-line basis reactor parameters including, flow, temperature, control rod position and pressure, (b) operating a digital computer to generate from said measured reactor parameters, nodal distributions of selected reactor parameters with at least one node per fuel assembly in the radial plane and a plurality of spaced nodes in the axial direction; to calculate from said selected nodal distributions, a stability index for selected fuel assemblies, taking into account core physical parameters which are a measure of power level, axial power distribution, flow, enthalpy, void drift, detailed fuel rod dynamics, nuclear reactivity feedback and, where appropriate, flow cross-coupling in the radial and axial directions; to determine the least stable selected fuel assembly from said calculated stability indexes; and to interatively recalculate the stability index for an unstable least stable selected fuel assembly using assumed incremental changes in a selected one of the rod position and flow parameters until said recalculated stability index for the least stable fuel assembly represents a stable condition; and (c) generating output representations of the calculated stability indexes and of the cumulative change in said selected parameter required to make the least stable fuel assembly stable; and (d) adjusting the reactor controls to change said selected one parameter by said cumulative change to bring said least stable fuel assembly to a stable condition. means for measuring on-line, selected reactor parameters including: flow, temperature, control rod position and pressure; a digital computer programmed: to identify in real time from the selected reactor parameters those fuel assemblies having inlet power peaking and above average power levels; to calculate only for the identified fuel assemblies a stability index taking into account hydrodynamic and nuclear feedback effects; and to generate output signals representative of said stability indexes; and means for generating human readable representations of said identified fuel assemblies and their stability indexes from said output signals. a reactor core having a plurality of fuel assemblies; control rods for adjusting the reactivity of said fuel assemblies in the reactor core; flow control means for controlling the flow of reactor coolant through the fuel assemblies of the reactor core; means for measuring selected reactor parameters including: flow, control rod position, temperature and pressure; digital computer means programmed: to generate nodal distributions of selected reactor parameters, with at least one node per fuel assembly in the radial plane and a plurality of spaced nodes in the axial direction; to select, using said nodal distributions, fuel assemblies having their peak axial power at a location below a selected location and having a power level above a preselected level to calculate from said nodal distributions a stability index for said selected fuel assemblies taking into account physical parameters which are a measure of power level, axial power distribution, flow, enthalpy, void drift, detailed fuel rod dynamics, nuclear reactivity feedback, and where appropriate, flow cross-coupling in the radial and axial directions; to determine the least stable selected fuel assembly from the calculated stability indexes; to iteratively recalculate the stability index for an unstable least stable selected fuel assembly using assumed incremental changes in a selected one of the rod position and flow parameters until said recalculated stability index for the least stable fuel assembly represents a stable condition; and to generate a control signal representative of the cumulative assumed changes in the selected one parameter; and control means to which said control signal is selectively applied to change the selected one of rod position or flow by the assumed cumulative amount. means for measuring on an on-line basis selected reactor parameters including flow, temperature, control rod position and pressure; a digital computer programmed to: (a) determine the radial and axial power distribution on a per fuel bundle basis using said selected parameters, (b) determine the peaking location for the axial power distribution for each fuel bundle and the average power level for the fuel bundles, (c) determine the stability margins only for those fuel bundles for which axial power peaking occurs below the average peaking location and for which the power level is above said average power level, and (d) iteratively repeat step (c) using incremental adjustments in a designated one of the flow and rod position parameters to determine a change in the designated parameter required to bring the stability margins for the selected fuel bundles within designated limits; and means for implementing said change in the designated parameter. 2. The method of claim 1 wherein said step of selecting fuel assemblies comprises determining the average axial location of axial power peaking and the average power for all of said fuel assemblies, and selecting those fuel assemblies with axial power peaking occurring below said average axial location and with power above said average power. 3. The method of claim 1 including: determining the least stable fuel assembly from said stability indexes calculated for said selected fuel assemblies, comparing the stability index of the least stable fuel assemblies to a preselected stability limit, and where said stability index for said least stable fuel assembly exceeds the preselected limit: 4. The method of claim 3 wherein said step of operating reactor controls to implement the cumulative change in the selected one parameter includes generating a control signal representative of the change required in the selected parameter and applying the control signal to an automatic controller to automatically effect the change in the selected one parameter. 5. The method of claim 3 wherein said monitoring step includes generating a real time power-flow map from said selected parameters and wherein said selected one parameter is flow, and said iterative calculations are carried out using incremental increases in flow and corresponding incremental increases in power determined from said power-flow map. 6. The method of claim 1 wherein said calculation of said stability indexes by said digital computer comprise nodal calculations with radially distributed nodes for each fuel assembly and a plurality of axially spaced nodes, and which incorporate nodal values of physical parameters which are a measure of nuclear feedback as well as the hydrodynamics of the flow channels. 7. A method of controlling the stability of the fuel assemblies in a boiling water reactor core comprising the steps of: 8. The method of claim 7 including operating said digital computer to select said fuel assemblies for which a stability index is calculated by determining for each fuel assembly, the power level, and from the axial power distributions, the location of the axial power peaking, and selecting those fuel assemblies with a power level above a preselected level and with axial power peaking occurring below a preselected axial location. 9. The method of claim 8 including operating said digital computer: to determine the average power level in said fuel assemblies and the average location of axial power peaking; and to select those fuel assemblies in which the power level is above said average power level, and in which axial power peaking occurs below said average location of axial power peaking. 10. A system for controlling instability in the fuel assembly of a boiling water reactor comprising: 11. The system of claim 10 wherein said digital computer is further programmed: to identify from the stability indexes the least stable fuel assembly; to iteratively recalculate the stability index for an unstable least stable fuel assembly using assumed incremental changes in a selected one of control rod position and flow parameters until the recalculated stability index for the least stable fuel assembly represents a stable condition; and to generate a control signal representative of the assumed cumulative change in the selected one parameter, said system including control means to which said control signal can be applied for changing the selected one parameter by the amount of the assumed cumulative change. 12. A boiling water reactor comprising: 13. Apparatus for controlling the stability of the fuel bundles in a boiling water reactor comprising:
	abstract	A subdermal lighting apparatus with enhanced operational longevity, safety and biological compatibility with a user is provided. The subdermal lighting apparatus includes a housing unit to store a radioactive isotope, a phosphor layer affixed to an interior surface of the housing unit to interact with radiation emitted by the radioactive isotope to generate visible light, a radiation shielding layer disposed around the housing unit and able to permit the generated visible light to pass through, and a biologically safe layer disposed around the radiation shielding layer able to permit the generated visible light to pass through. The biologically safe layer serves as a barrier between the radiation shielding layer and biological tissue of the user, thereby enhancing user safety of the apparatus.
	abstract	A method for performance monitoring in a computing system is described. In some embodiments, an addressable memory stores data and instructions for performing context switch sampling. A processor includes hardware event counters, and is coupled with the addressable memory to access said instructions and in response to said instructions, the processor counts occurrences of a first hardware event in a first hardware event counter and counts occurrences of a second hardware event in a second hardware event counter. After a specified number of occurrences of the first hardware event have been counted, the second hardware event counter is sampled and hardware event counters are reset. In some embodiments the processor counts occurrences of segment register load events in the first hardware event counter and then records the sampled second hardware event counter value with a process identifier value and/or a thread identifier value.
041918886	summary	BACKGROUND OF THE INVENTION The present invention generally relates to an ion extraction system for devices utilizing grids to accelerate and focus ion beams, such as ion thrusters and ion sources. More particularly, this invention utilizes small hole accel grid (SHAG) optics, which minimizes the emission of neutral atoms with a three-electrode system and allows high ion extraction current densities at low net extraction voltages. A particular feature of the invention is the construction of the accel grid that has the advantages of reducing the contamination of surfaces downstream of the electrode system through accel grid sputtering and is easier to fabricate and use than a conventional SHAG electrode. Present ion extraction systems use two or three electrode systems with either standard hole accel grids, having an open area approximately 50% that of the aperture of the screen grid, or small hole accel grids, having an open area as low as 10% of the aperture of the screen grid. In the latter case, the hole sizes are slightly larger than the individual ion beamlets. The state-of-the-art system utilizes small hole accel grid (SHAG) optics with a three-electrode system comprising a screen grid, the accel grid and a decel grid. A neutral highly ionized plasma upstream of the electrodes is typically held at a high positive voltage of several hundred to several thousand volts. On the other hand, the accel grid is held at a negative voltage of several hundred to several thousand volts, while the decel grid is held at zero or ground potential. The electric field set up by the screen grid and the accel grid extracts an ion beam from the plasma source and focuses it through the accel grid hole. The function of the decel grid is to minimize the beam spread or divergence after the ion beam passes through the accel grid. The small aperture size of the accel grid minimizes the emission of the un-ionized atoms from the discharge plasma, which in ion thrusters maximizes the thruster performance, and in ion sources minimizes the propellant gas flow, thus minimizing the vacuum system pressure. One disadvantage of this electrode system is the possibility of contamination by sputtered accel grid material. As the ion beam passes through the accel grid aperture, charge-exchange between the high energy ions and the neutral atoms drifting through the aperture can occur, producing a low energy ion which strikes the accel grid. In a three-electrode system, these charge-exchange ions strike and sputter the grid on the barrels of the holes. Although most of the sputtered grid material will be contained within the grids and discharge chamber, some of the material is emitted and can coat surfaces downstream of the grids. Another shortcoming of the electrode system is the difficulty in making the SHAG electrodes. The usual method has been to use a blank plate at the accel grid and let the ion beamlets etch each hole. This process is fairly tedious and can take 24 hours or longer with fairly thick accel electrodes. If the electrode assembly is disassembled and reassembled, perfect realignment is practically impossible so that the electrode system will have to be operated for a time before the ion impingement on the accel grid is reduced to a negligible value. SUMMARY OF THE INVENTION The present invention overcomes both of the deficiencies over the prior art electrode system described above. According to the invention, the accel grid has a large, drilled aperture, typically of a diameter on the order of 2.083 mm. A thin foil, for example, tantalum of a thickness of 0.0127 mm is placed on at least the downstream face of the accel grid and covers the large, drilled aperture. In the preferred embodiment, a second very thin metallic foil is also placed on the upstream face of the accel grid covering the large, drilled hole. With this construction, the ion beam etches only through the thin foils so that the etching time is much shorter than for a solid plate, sometimes less than an hour with very thin foils. Sputtered acceleration grid material contamination is substantially reduced because the doughnut-shaped volume defined by the interior foil surfaces and the barrel of the drilled hole traps much of the sputtered material. If the electrode system is disassembled and reassembled, the run-in time before direct accel grid impingement disappears is much less than for a solid plate SHAG electrode because only the thin foils have to be etched.
046722114	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to blackbody simulators, and more particularly to such simulators having a core with an aperture to an interior cavity, the aperture simulating the properties of a similarly sized and shaped blackbody. 2. Description of the Prior Art A blackbody is an idealized object which would absorb all electromagnetic radiation impacting it. Since such an object wouls absorb all light striking it, it would appear black. The physical properties of blackbodies have been intensively studied. Although the definition of a blackbody is in terms of its perfect absorption of electromagnetic radiation, its most interesting properties are associated with its radiated energy. When considered as a radiation source, it is generally considered to be heated to increase its radiated energy. The total emission of radiant energy from a blackbody is expressed by the Stefan-Boltzmann law, which states that the total electromagnetic emission of a blackbody is proportional to the fourth power of its absolute temperature. The spectral energy distribution of the radiant energy emitted by a blackbody is expressed by Planck's radiation formula. Planck's radiation formula indicates that a blackbody which has a temperature between 50 degrees Kelvin and 3,000 degrees Kelvin will emit electromagnetic radiation principally in the infrared region. This temperature range encompasses the temperatures at which most nonnuclear physical phenomena occur. A blackbody is an idealized concept. A blackbody simulator is an apparatus designed to simulate the physical properties of the idealized blackbody. A blackbody simulator is of great use in infrared research and development and manufacturing. For instance, it may be used to provide a source of infrared radiation of a known signal level and a known spectral distribution. It may be used to provide a source of infrared radiation for the adjustment or testing of infrared components, assemblies or systems. Another use for blackbody simulators is as a near perfect absorber of infrared radiation. Intuitively, it would appear that a blackbody simulator would be merely any black object. Such simulators have been used in the past, but their correspondence to a true blackbody has been poor. The best blackbody simulators are formed by creating a cavity in a core material, the cavity forming an aperture on a side of the core. The aperture is used to simulate a flat blackbody having the shape and size of the aperture. Particular cavity shapes are chosen to cause multiple reflection and eventual absorption of any electromagnetic energy entering the aperture. One measure of the closeness by which a blackbody simulator approaches a true blackbody is its emissivity. The emissivity of an isothermal surface is the ratio of the radiation emitted by the surface to the radiation emitted under identical conditions by a blackbody having the same shaped surface and temperature. Unless the blackbody simulator is luminescent, the emissivity of an isothermal blackbody simulator is less than one. If the cavity surface coating of a blackbody simulator has an emissivity above 0.7, most cavity shapes in commercial use in blackbody simulators will result in a device in which the on-axis emissivity of the aperture will exceed 0.99. The angular distribution of radiation from a perfectly diffuse radiator, such as a blackbody, is given by the Lambert cosine law. This law states that the signal level of radiation from a perfectly diffuse surface is proportional to the cosine of the angle between the direction of emission and a normal to the surface. Accordingly, the Lambert cosine law provides a standard against which the uniformity of emission from a blackbody simulator can be compared against a blackbody. It is well known how to manufacture a blackbody simulator with a cavity shape configured to have an on-axis emissivity very close to one. There are commerically available blackbody simulators with an on-axis emissivity of 0.9997. Unfortunately, the prior art cavity shapes for high emissivity blackbody simulators have been found to not provide the uniformity of emission specified by the Lambert cosine law. Such uniformity of emissivity is becoming a significantly more important consideration in current infrared research and development. It appears that the uniformity of blackbody simulator emissivity will become a significantly more important consideration in future manufacturing adjustments and tests of infrared components, assemblies and systems. For instance, infrared viewing systems have been the subject of intensive research and development. Initially, a single element infrared sensor was used with a two axis mechanical scanner to provide a two dimensional infrared picture. For calibration and testing of such single elements, a blackbody simulator was used to illuminate a relatively narrow field of view. Since a single element infrared sensor was used, uniformity of emissivity was not as important as it is today. However, infrared viewing systems now in development use a one dimensional or two dimensional array of infrared sensors, thereby eliminating the need for one direction or both directions of mechanical scanning. Testing and calibration of such array detector systems require a blackbody simulator which has an emissivity which is uniform over the appropriate angular field of view. Although emissivity and uniformity of emissivity are important specifications for a blackbody simulator, such simulators also are developed with other very practical considerations in mind. Key specifications for a blackbody simulator to be used as a source of infrared emission include aperture size and temperature range. The cost of a black body simulator is in large part determined by the physical size and weight of the blackbody simulator. The cost is also affected by the particular cavity shape used inasmuch as certain cavity shapes are more expensive to manufacture. The radiation properties of a cavity type blackbody simulator are determined by the size and shape of the cavity and the temperature, material and texture of the cavity walls. Most commercially available blackbody simulators have a cavity shape based upon a cone. Other popular shapes are the sphere, the reentrant cone and the cylinder. A reentrant cone cavity has a shape which is formed by placing base to base two circular cones having the same size base, and truncating the apex of one of the cones to form the aperture. Both a conical cavity and reentrant cone cavity have a conical apex opposite the aperture. An advantage to a blackbody simulator having a spherical cavity is that its emissivity is very uniform. Further, it has been theoretically proved and experimentially verified that the surface of a spherical cavity tends to become isothermal, i.e., the cavity surface temperature tends to become uniform. It is desirable for the surface of the cavity to be as nearly as possible isothermal since that property is necessary for the spherical distribution of the simulator to conform to the Planck radiation formula. Despite these advantages, a spherical blackbody simulator is a large object, and is correspondingly very heavy. In addition, its on-axis emissitivity can be achieved by considerably smaller blackbody simulators having conical or reentrant cone cavities. Another disadvantage to a blackbody simulator using a spherical cavity is specular reflection from the cavity wall opposite the aperture. Radiation entering the aperture on-axis tends to be reflected out the aperture, rather than be absorbed. Such specular reflections are antithetical to the definition of a blackbody. This latter deficiency of a spherical cavity blackbody simulator is sometimes overcome by using a tilted or off center arrangement for the sphere and the core, but this technique requires even larger and heavier assemblies. Blackbody simulators having a generally cylindrical cavity shape, with the aperture in one axial end of the cylinder, are easy to manufacture. Some cylindrical cavities have been provided a concentrically grooved back plate where the grooves have a generally uniform saw tooth cross-section, such as that shown in De Bell, U.S. Pat. No. 3,419,709. Unfortunately, cylindrically shaped blackbody simulators have lower on-axis emissivity than a cone, and they have poor uniformity of temperature and poor uniformity of emissivity. Most commercial blackbody simulators have a cavity shape which is either a cone (such as that shown in McClune, et al., U.S. Pat. No. 3,275,829) or reentrant cone (such as that shown in Stein, et al., U.S. Pat. No. 3,699,343). Such shapes have excellent on-axis emissivity. Specifically, the apex of the cone opposite the aperture has emissivity much greater than that of a sphere with a diameter such that the wall of the sphere opposite the aperture would be at the same distance from the aperture as that of the apex of a cone or a reentrant cone. A blackbody simulator having a cone as a cavity shape is relatively easy to manufacture, except for the apex, but it has poor uniformity of emission. The reentrant cone is more difficult to manufacture, but the advantage of a reentrant cone over a simple cone is that the surface of the cavity maintains a more uniform temperature and emissity than that of a similar sized cone, since a reentrant cone cavity tends to minimize the cooling of the cavity near the aperture. Parallel v-groove, circular v-groove and honeycomb arrays of hexagonal cross-section tubes have been used in recent years. These shapes are most often associated with large area blackbody simulators. For most of these designs the projected solid angle of the apertures or forward facing openings as seen from various points on the cavity wall surface vary from rather small values for points that are deep in the grooves or far from the apertures (or forward facing openings) to larger or rather large values for points that are near the apertures (or forward facing openings). It is an object of the invention to provide a cavity type blackbody simulator that has the high on-axis emissivity obtained with cone or reentrant cone blackbody simulators, yet also approaching the uniform emissivity obtained with spherical blackbody simulators. Another object of the invention is to provide a compact blackbody simulator which is shorter and smaller in diameter than a blackbody simulator having a spherical cavity with the same emissivity. Yet another object of the invention is to provide a blackbody simulator with a cavity having a shape which tends to take a uniform temperature when heated. A further object of the invention is to provide a method for the design of a blackbody simulator cavity having a specified maximum depth, diameter, and aperture size in which high emissivity and uniformity of emissivity are obtained. SUMMARY OF THE INVENTION These and other objects of the herein disclosed invention are provided by a blackbody simulator having a cavity shaped to insure that the value of the projected solid angle of the aperture is constant or approximately constant when viewed from all points on the cavity surface, or alternately, when viewed from all points on the primary radiating surface of the cavity. The primary radiating surface of a blackbody simulator cavity with respect to an object outside the cavity is the portion of the cavity surface which is in direct line of sight to some portion of the object. The projected solid angle of one surface as viewed from a second surface is proportional to the fraction of the total amount of radiant energy leaving the second surface that impinges on the first surface. In the embodiments of the present invention, numerical integration techniques have been used to determine a series of cavity shapes which maintain essentially constant the projected solid angle of the aperture with respect to each portion of the cavity surface or at least the primary radiating surface. The cavity shape is formed by smoothly arcuate subsurfaces. The first subsurface forms a cone-like apex opposite the aperture, the subsurface smoothly curving a predetermined distance toward the aperture. The second subsurface starts from the rim of the aperture and proceeds toward the apex; it is a portion of the surface of a sphere. One or more additional subsurfaces arcuately join the two previously described subsurfaces. The subsurfaces of the primary radiating surface are determined so that the projected solid angle of the aperture as seen from all points on all these subsurfaces of the primary radiating surface is constant. In accordance with the present invention, the structure of the above-described primary radiating surface may be simplified by increasing the apex angle for certain cavity shapes. Moreover, the size, weight and cost of the simulator may be further reduced by decreasing the ratio of the cavity depth (L) to the aperture diameter (D).
	abstract	Apparatuses, processes and methods for the separation, isolation, or removal of specific radioactive isotopes from liquid radioactive waste, these processes and methods employing isotope-specific media (ISM). In some embodiments, the processes and methods further include the vitrification of the separated isotopes, generally with the ISM; this isotope-specific vitrification (ISV) is often a step in a larger scheme of preparing the radioactive isotopes for long-term storage or other disposition. A variety of ISM are disclosed.
	abstract	A method includes sterilizing a column assembly including a column having a parent radionuclide contained therein with a sterilizer. The method further includes transferring the column assembly from the sterilizer to a first clean room environment, transferring the column assembly from the first clean room environment to a second clean room environment, and collecting a sterility test sample from the column assembly within the second clean room environment.
051006090	summary	BACKGROUND OF THE INVENTION The present invention relates to nuclear boiling water reactors (BWRs) that utilize natural circulation and more particularly to enabling load-following capability and/or enhancing spectral shift in such reactor designs. Existing large BWRs are of the forced-circulation type. In BWRs undergoing power generation operations, reactor coolant, initially in the form of sub-cooled liquid (e.g. water), is circulated by main coolant recirculation devices (e.g. jet pumps or mixed-flow motor-driven pumps) around a path a portion of which is comprised of the core lower plenum region (located at the bottommost section of the reactor), thence through the nuclear core and into a core upper plenum in communication with the core. Flow exiting the core upper plenum then passes through standpipes that lead to an assembly of steam separators. The reactor coolant exiting the nuclear core and passing into the core upper plenum is a two-phase mixture of steam and water, the proportion of which varies depending upon such factors as the power output from the fuel bundles, the amount of sub-cooling present in the coolant entering the fuel bundle, and the amount of flow through the bundles. This last factor depends on the power of the recirculation pumps and the hydrodynamic flow resistance presented by the fuel bundle geometry and wetted surfaces, and the amount of orificing representing restrictions to flow just prior to the coolant's entrance into the core fuel assembly. Joining with the core effluent in the core upper plenum is the core "by-pass" flow, which is reactor coolant that has flowed from the core lower (entrance) plenum into the region external to the fuel assembly channels (but inside the core shroud), thence upwardly generally through the region occupied by cruciform-shaped control blades which stand in various degrees of insertion into the core, thence across the upper grid member (called the "top guide") which with its lattice-like configuration keeps the fuel assemblies in a regular array, and finally into the core upper plenum. This by-pass coolant stream at its discharge into the core upper plenum is compromised substantially of saturated liquid, with perhaps a small amount of steam. Within the core upper plenum, these two effluents-by-pass flow and fuel bundle exit flow-rapidly mix together and quickly lose identity from their origins. Mechanical steam separation can be utilized to accomplish the separation of the steam from the steam/water mixture exiting the core. Some earlier BWR designs used free-surface steam separation where, just as in the household tea kettle, steam separates unaided from the free-surface, and saturated water remains in the bulk coolant, which in BWRs is recirculated back down the downcomer annulus. This type of steam separation is feasible so long as the steam-leaving velocity, i.e. the bulk average velocity of the steam taken across the available pathway flow area, is not large, i.e. is no greater than about 1.8 foot/second. If steam-leaving velocities exceed this value, there tends to be carried along with the steam an unacceptably high moisture content. The high moisture levels saturate the moisture-drying abilities of the steam dryer, thus resulting in an unacceptably high moisture content in the steam leaving the reactor and supplied to the turbine. When steam moisture contents are too high in the turbine steam flow, accelerated erosion can occur on first-stage turbine blades and the efficiency of the turbine is reduced. It is possible to obtain free-surface separation capabilities if the reactor pressure vessel (RPV) cross-sectional area is made sufficiently large. However, cost economies dictate that minimum diameter RPVs be used, so that mechanical steam separation has been developed to handle the high power output steam production levels of modern BWRs. In these latter designs, the steam bulk average velocity moving through the wet steam plenum region immediately downstream of the mechanical steam separators is about 5 feet/second. The fuel assemblies grouped over the central region of the core tend to have higher exit steam qualities than do bundles located at the peripheral region of the core. It is desirable, nonetheless, that the flow rates and steam/water mixture proportions entering the steam separator standpipes be relatively uniform. To facilitate gaining more nearly uniform steam/water mixture for entry into the standpipes, the standpipe entrances are separated from the fuel assemblies by a distance of, for example, about 5 feet. Turbulent mixing occurring between the plumes leaving adjacent fuel assemblies, each with a different void content, is one mechanism acting to produce a more nearly uniform mixture which enters into the steam separator standpipes. More important to achieving flow mixture uniformity, however, is the hydrodynamic flow resistance represented by the standpipes, each with their end-mounted steam separators. Complete flow mixture uniformity entering the standpipes is at best difficult to achieve and, even with a five-foot separation between fuel assembly exits and standpipe entrances, it is not a design basis used for reactor performance evaluations. The steam separator assembly consists of a domed or flat-head base on top of which is welded an array of standpipes with a three-stage steam separator, for example, located at the top of each standpipe. One function of the standpipes is to provide a stand-off separation of the larger-diameter steam separators, which are generally arranged in a particularly tightly-compacted arrangement in which external diameters of adjacent separators are nearly touching with each other, so that separated liquid coolant discharged at the bottom of the separator has a more "open" flow path outwardly from the reactor longitudinal axis and out to the downcomer annulus region which lies at the inboard periphery to the RPV. A second purpose for the standpipes is a high-power-output natural-circulation reactor using mechanical steam separators is to provide juxtaposed regions which promote natural-circulation by means of a vertical region of two-phase (and, thus, low-density) coolant inside the standpipes which is juxtaposed against single-phase liquid coolant outside the standpipes in a so-called "downcomer region", in which region height provides a very significant part of the total natural circulation driving head for coolant flow circulation within the reactor. The steam separator assembly rests on the top flange of the core shroud and forms the cover of the core discharge plenum ("core upper plenum") region. The seal between the separator assembly and core shroud flange is a metal-to-metal contact and does not require a gasket or other replacement sealing devices. The fixed axial flow type steam separators have no moving parts and are made of stainless steel, for example, to resist corrosion and erosion. In each separator, the steam/water mixture rising through the standpipes (the "standpipe region") impinges upon vanes which give the mixture a spin, thus enabling a vortex wherein the centrifugal forces separate the water from the steam in each of three stages. Steam leaves the separator at the top of this assembly and passes into the wet steam plenum below the dryer. The separated water exits from the lower end of each stage of the separator and enters the pool (the "downcomer region") that surrounds the standpipes to join the downcomer flow. The steam exiting from all separators either may be in the same horizontal plane, or the separators may be arranged in a slightly crowned fashion at the center to compensate for the crowned water gradient of the pool surrounding the standpipes. The steam separator assembly may be bolted to the core shroud flange by long hold-down bolts, or the separator together with the dryer assembly may be held down onto the core shroud flange by contact from the reactor head when the latter is assembled to the reactor vessel. The nominal volumetric envelope of the steam separator assembly is defined by the horizontal plane of its lower flange that contacts the core shroud flange, its cylindrical sides that provide part of the five-foot stand-off from the fuel assembly exits, the circumscribed diameter of the outermost row of standpipes, the circumscribed diameter of the outermost row of steam separators, and the generally horizontal plane of the exits to the steam separators. The core upper plenum region in a BWR currently under design known as the "simplified boiling water reactor" (SBWR) is substantially devoid of other mechanical devices, pipes, or structures; whereas the core upper plenum of a BWR/6 and "advanced boiling water reactor" (ABWR) reactor design generally contains spargers and nozzles for core sprays, and distribution headers for core flooders, respectively. In both reactor types, these spargers/headers are located at the outer periphery of the core upper plenum, mounted below the core shroud flange so that the sparger/header is clear of the refueling removal path of peripheral fuel assemblies and, thus, do not become removed during core refueling operations. With specific reference to a natural-circulation SBWR, it will be observed that there are no recirculation pumps to aid in coolant recirculation. Steam generation in the core produces a mixture of steam and water which, because of steam voids, is less dense than saturated or sub-cooled water. Thus, the boiling action in the core results in buoyancy forces which induce core coolant to rise upwardly, to be continuously replaced by non-voided coolant arriving from beneath the core in the core lower plenum region. As the coolant leaves the core, it rises through the core upper plenum region, then through the standpipes region, and finally into the steam separators. This voided mixture inside these standpipes continues to be less dense than non-voided coolant external to the standpipes, resulting in the development of additional buoyancy force to further drive the coolant circulation. That this process is quite effective in promoting coolant recirculation can be noted from reported tests made in forced-circulation power reactors where the coolant circulation pumps are shut off. Even with their relatively short steam separator standpipes, reactor power levels of 25% and coolant flow rates of 35% of rated flow, are readily and safely maintainable. The SBWR reactor is but modestly different from the forced-circulation BWR, with the most prominent differences being that the standpipes region is to be considerably longer in the SBWR (to develop a higher differential head), the core overall height may be somewhat shorter (for example, being 8 or 9 feet active fuel length versus 12.5 feet active fuel length in recent forced-circulation reactors), and the core power density will be somewhat lower. The severity of orificing--a means to promote hydrodynamic stability--at the entrance to the BWR fuel bundles may be lessened. The fuel bundle may have a larger diameter fuel rod in, for example, a 6.times.6 rod array, whereas the rod array for a forced-circulation reactor often is an 8.times.8 rod array. The design flow rates per fuel bundle, and the flow rates per steam separator, will be somewhat reduced in the SBWR design. Fuel exit steam quality will be approximately the same between the two designs. In the SBWR reactor design, no spargers or discharge headers are installed in the core upper plenum, while in the ABWR reactor, spargers or discharge headers are installed in the upper core plenum. In some versions of SBWR reactors under study, the standpipes are very long while the core upper plenum is short. In other versions, the converse is true. The present invention is applicable equally in either version. "Load-following" is the action of bringing the power output of a BWR into balance with an incrementally changed power output demand. This demand change arises from the electrical grid to which the nuclear power station is coupled and represents a change from prior steady-state (balanced) operating conditions. By way of illustration, assume that an SBWR is operating at 90% of rated power output. Existing within the core will be some certain distribution of voids, i.e. steam vapor in the form of steam bubbles. The lowermost parts of the fuel assemblies will contain nonvoided coolant because of the sub-cooled liquid conditions existing in the core lower plenum, the source for water entering the core. Partway up the flow path within the fuel assemblies, steam generation begins, so that a steam/liquid mixture develops with the steam proportion rising with increasing travel upward through the fuel assembly. Control blades immediately outside the fuel assembly channels will stand in various degrees of withdrawal from the core depending on the particular point the core has reached in its fuel cycle lifetime. The steam output from the nuclear boiler is coupled to a turbine generator which, in turn, is coupled electrically to the grid. A nuclear boiler pressure regulation control system is installed, the action of which changes the position of turbine steam control valves in such a way so as to maintain constant the nuclear boiler pressure as measured in the reactor steam dome. A change in grid electrical demand--say an incrementally increased demand for more electrical power from the power station--causes a signal to be sent to a control rod positioning system that results in incremental withdrawal of certain of the control blades still not fully withdrawn from the core. This withdrawal has the effect of making the reactor temporarily more reactive, allowing an increase in neutron flux, that, in turn, produces a higher rate of nuclear fission throughout the fuel rods. The thermal capacitance represented by the mass of the fuel material (uranium dioxide) briefly, i.e. for a few seconds, absorbs the thermal energy produced throughout the fuel rods as their internal temperature rises. (The fuel heat transfer thus lags the neutron flux, the transient response characteristic being that of, typically, a seven-second time constant). Soon, however, the higher temperatures lead to greater heat transfer from the now-higher fuel clad temperature to the reactor coolant, and so an incrementally higher amount of steam is formed. In addition, the point where boiling first begins within the fuel assembly moves slightly downward in response to the higher heat transfer that is occurring. This combination of incrementally more voids in prior boiling regions, plus downward movement of the boiling boundary, now introduce negative reactivity effects that returns the reactor to a balanced, steady-state, power level, but one that is generating incrementally more steam. In response to the larger steam generation rate, to hold pressure constant in the reactor steam dome (as the control system mandates), the pressure regulation system progressively incrementally opens the turbine control valves, thus releasing a net greater quantity of steam to the turbine. Higher steam rates passing through the turbine produce the required incremental increase in ultimate response to the initiating electrical grid demand for more electrical power from the station. The principles described above can be extended to those conversant with nuclear engineering practice to understand other types of power adjustments. It will be apparent that the foregoing illustration is the response to a small increase in load demand. Obviously, the adjustment of reactor power output also can be performed manually by the reactor operator, through his actions to cause control blades to be inserted farther into the core or withdrawn farther outward from the core. Often, a nuclear power station is required to sustain larger load demand adjustments than the relatively small adjustment described above. Existing nuclear power stations are deficient in that it takes time to retract the control blades. Even when the control blades are moved in groups ("ganged rod movement"), it still requires time for the groups to be sequentially moved. An additional disadvantage to load-following by control blade movement can be that the heating transients within the fuel occurring close to the ends of where the control blades are positioned, over time can produce undesirable stress-cycling on fuel cladding. An alternative load regulation means that has been found effective for forced-circulation BWRs is to use recirculation flow control. A signalled change in reactor power demand is sent to a control system that adjusts to recirculation flow upward or downward. The recirculation flow is regulated either by changing the speed of the main recirculation pumps, or in other applications by throttling the output; from constant-speed pumps by means of a flow-control valve. The changed flow causes a rather prompt change in the amount of voids in the core and a similar change in the position of the boiling boundary within the fuel assemblies within the core. For example, the action response in the recirculation flow control system to an incremental demand for more reactor output would be to raise the rate of recirculation flow. This sweeps some existing voids out of the core, and raises the position of the boiling boundary. In turn, neutron flux rises, fission rate increases, and shortly a higher total amount of steam is being regenerated. With the reappearance of "near-normal" levels of voids in the core in response to the higher power output, the reactor condition returns to a "steady-state", but now at a higher output level. The two advantages of recirculation flow control are the rates of change in reactor core power can be faster; and since control blades are not required to be moved, no additional significant stress-cycling duty is imposed on the fuel rods. To date, however, natural-circulation reactors have had only control rod movement available to them as a load-following means. As described earlier, a drawback with this mode of load-following control, i.e. performing load-following by moving control blades, is that it can be a slow-acting system because there are so many blades which just be moved some variable small amounts to effect a change while keeping the neutron flux profiles in desirable patterns. Thus, for various modes of power operation, it would be desirable for there to be other methods of more rapidly, yet controllably, affecting reactor power output, and thereby provide enhanced capability to perform a wider envelope of load-following maneuvers. Another operation that can be accomplished in forced-circulation type BWRs is known as "spectral shift". "Spectral shift" of neutrons is a shift in the energy level of neutrons in the reactor core that enables a non-fissile material to be transmuted into a fissile material. A typical enhancement involves the transmutation of .sup.238 U to .sup.239 Pu. Those skilled in the art already are aware of the fact that most sources of mined uranium are subjected to a variety of operations including concentration, conversion, and enrichment, in order to supply fissile material that is provided in the form of fuel elements for use in forming fuel rods for use in nuclear power plants, for example. Depending upon the process utilized and the type of reactor involved, cost and/or technological considerations result in a defined amount of non-fissile material yet being present in the nuclear fuel. By transmuting the non-fissile fraction of the fuel to a fissile form, the nuclear reactor can be operated for a longer period of time without refueling and/or before fewer fresh fuel bundles need to be installed. Thus, spectral shift can be a desirable mode of operation for various reactors. In forced circulation BWRs, spectral shift is achieved by reducing the recirculation rate which causes an increase in the void fraction in the core. Power level is reduced thereby. Accordingly, control rods are withdrawn to re-establish the reactor power level desired. When the void fraction is higher, the neutron spectrum shifts to a higher energy level, thus causing the non-fissile material to be transmuted to fissile form. Since natural-circulation reactors operate without forced water circulation, other techniques need to be developed in order to operate the reactors in a spectral shift mode. BROAD STATEMENT OF THE INVENTION The present invention discloses an improved methodology for achieving the desirable feature of recirculation flow control in a sparger-type natural-circulation or forced circulation BWR, enabling reactors utilizing the method to perform in load-following and/or spectral shift modes of power operations. Accordingly, the present invention is directed to a natural-circulation and forced circulation nuclear boiling water reactor of the type wherein a reactor pressure vessel (RPV) contains a lower water inlet and upper steam outlet connected to an upper steam dome. Housed within the RPV is a nuclear core disposed within a shroud which with the RPV defines an annulus region which is in flow communication with a core lower plenum region disposed beneath the nuclear core. A core upper plenum formed within the head of the shroud is disposed atop the nuclear core. The sparger is located above the core upper plenum. The BWR may be of the type utilizing natural steam/water separation wherein steam separates unaided from the free-surface of the water housed within the chimney of the RPV. Alternatively, the BWR may be of the type wherein a plurality of vertically-mounted standpipes are mounted atop the shroud head (and comprising the chimney) and in fluid communication with the core upper plenum and the upper steam dome. A steam separator/dryer assembly optionally may be mounted atop the standpipes and in fluid communication with the standpipes and with the chimney. The chimney is in fluid communication with the downcomer region located outside of the chimney (e.g. standpipes) to complete the water flow path. A water level is established within the BWR under normal (a defined power output level) operation of the BWR. The improvement comprises the RPV feedwater inlet being connected to a sparger that is located about the interior circumference of the RPV. The sparger is located at an elevation at about the liquid water level established during normal operation of the BWR. This elevational location of the feedwater sparger enables one or more of load following or spectral shift to be performed by controlling the feedwater fed to the BWR through the feedwater inlet and into the sparger to vary the water level to be above, at, or below the elevational location of the sparger. Varying the feedwater fed to the BWR, and consequently raising or reducing the water level, reduces or increases the amount of steam that is condensed and increases or reduces the amount of sub-cooling and, consequently, the reactor power. Alternative configurations of common reactor internals with modifications and/or additional components are described as illustrations of the application of the invention to reactor design. Advantages of the present invention include a new means for providing load-following capability for sparger-type natural-circulation BWRs. A further advantage of the present invention is a new means for enhancing spectral shift capability for sparger-type natural-circulation BWRs. Yet a further advantage of the present invention includes a reactor design that is capable of being operated in a load-following mode, a spectral shift mode, or a combination of these modes. Another advantage of the present invention is the ability to provide adjustable regulation of the feedwater to the reactor to effect coolant recirculation flow rate. These and other advantages will be readily apparent to those skilled in the art based upon the disclosure contained herein.
049838512	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relate to a contact therapeutical apparatus for giving physical treatments such as magnetic and heat therapies by contacting a human body. 2. Technical Background I have, in the past, proposed a magnetic field generating therapeutical apparatus constructed by connecting a plurality of sealed cases containing a magnetic field generator consisting of an iron core wound with coils. The apparatus is so designed that when an AC current is applied to the coil, the magnetic field generator generates an alternating field which is applied to the body to improve various symptoms such as stiffness and so on. In the magnetic field generating therapeutical apparatus, the magnetic field generator generates vibration and heat when energized, thus in addition to the magnetic therapeutical effect, massaging and warming effects can be brought to a body. In the case of aforesaid the apparatus, however, since heat generated in the magnetic field generator functions mainly on the surface of the body and lacking in warming the inside thereof, the warming effect is not sufficient. The present invention has been devised in view of the problems aforementioned, and therefore, it is an object thereof to provide a novel contact therapeutical apparatus having a superb warming effect by providing on the case body a heat generating portion or a cover member covering the case body. It is another object of the present invention to provide a novel contact therapeutical apparatus capable for warming the body uniformly to give the warming effect more effectively and provide a comfortable feeling. DISCLOSURE OF THE INVENTION The present invention relates to a contact therapeutical apparatus having a heat generating portion in a case body contacting the body, and containing a far infrared radiating material in the case body or the cover member covering the case body. According to the present invention, when heat is generated in the heat generating portion, it is conducted to the case body or cover member, thereby the far infrared radiating material is excited to radiate far infrared rays. The far infrared rays radiated is applied to the body to warm inside the body to give effective warming effects. Since the far infrared radiating material is heated by the heat generated from the heat generating portion of the apparatus and radiates the far infrared rays, a special heat source is not required, resulting in simple production at low cost. In the present invention, on the side of the case body contacting the body, for distributing heat uniformly, there is provided a uniformalizing layer, or inside the cover member covering the case body, the same uniformalizing layer is provided and at the same time, the far infrared radiating material is contained in the uniformalizing layer. According to the present invention, heat is distributed uniformly to warm the body evenly in a wide area, thereby a more effective warming effect and comfortable feeling can be obtained.
	summary
	abstract	The present invention provides a method adhering a layer to a substrate that features defining first and second interfaces by having a composition present between the layer and the substrate that forms covalent bonds to the layer and adheres to the substrate employing one or more of covalent bonds, ionic bonds and Van der Waals forces. In this manner, the strength of the adhering force of the layer to the composition is assured to be stronger than the adhering force of the layer to the composition formed from a predetermined adhering mechanism, i.e., an adhering mechanism that does not include covalent bonding.
	summary
	abstract	A method for encapsulating a fuel rod or a fuel rod section in a container includes inserting the fuel rod or fuel rod section into the container. One of the ends of the container is connected to a purging-gas line. The container is dehydrated and purged by use of a purging gas. The ends of the container are connected to a bypass line in such a way that a closed gas circuit is produced and a hot gas is circulated in the gas circuit until the absolute moisture content reaches an end value at which the absolute moisture content no longer rises. The container is disconnected from the gas circuit and subsequently the container is closed in a fluid-tight manner at both ends.
	abstract	The invention relates to an optimized laser cutting method for cutting a part from a material by means of a cutting system comprising: a laser source for producing a laser beam having a certain power; and a cutting head comprising an end nozzle for passage of a cutting laser beam, said method being characterized in that it comprises a step of determining the cutting power Pd such that: Pd=Max(Popt;λe) where Max is the mathematical operator of the maximum, Popt is an optimal power of the laser beam of the cutting system, which is predetermined in accordance with the part to be cut, and/or with cutting parameters and/or with system parameters, to minimize the mass defect per unit length of the part when the part is being cut, λ is a leading coefficient representing the number of kW required for cutting the part per mm of the thickness of the part, and e is the thickness of the part in mm.

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