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	claims	1. A backscatter-based X-ray staring imaging apparatus comprising:at least one X-ray generator for irradiating a selected target with a non-scanning beam;an X-ray focusing device positioned for receiving X-ray backscatter from said irradiated target; anda detector positioned relative to said focusing device for forming an image from said received X-ray backscatter;wherein said focusing device comprises a lobster eye structure;wherein said lobster eye structure comprises a decussate arrangement of interleaved wafer-thin flat ribs configured as a unitary three-dimensional meshed array forming a plurality of contiguous channels of continuously diminishing square cross-section, each of said flat ribs being a segment of an annulus having an inner circumference and an outer circumference. 2. The imaging apparatus recited in claim 1 wherein each of said flat ribs has a polished reflective surface and a plurality of spaced mating slots for being interleaved with other flat ribs. 3. The imaging apparatus recited in claim 1 wherein said detector comprises a digital device for converting said focused X-ray backscatter into an electronic image. 4. The imaging apparatus recited in claim 3 wherein said digital device comprises a scintillating screen coupled to a COD matrix by a fiber optic taper. 5. The imaging apparatus recited in claim 4 wherein said scintillating screen comprises a microchannel plate having a plurality of microchannels filled with a scintillating material. 6. An X-ray inspection system for generating an image of at least one object hidden behind a wall; the system comprising:at least one X-ray source generating open cone X-ray radiation for irradiating an area to be inspected which area includes said wall and object;a digital X-ray imaging detector for converting an X-ray image into an electronic image;a lobster eye-based X-ray focusing structure for receiving and focusing X-ray radiation backscattered from said inspected area and onto said imaging detector for imaging said at least one object;wherein said lobster eye-based X-ray focusing structure comprises a decussate arrangement of interleaved wafer-thin flat ribs configured as a unitary three-dimensional meshed array forming a plurality of contiguous channels of continuously diminishing square cross-section, each of said flat ribs being a segment of an annulus having an inner circumference and an outer circumference. 7. The X-ray inspection system recited in claim 6 wherein said X-ray source open cone radiation is a non-scanning, staring-type X-ray beam. 8. The X-ray inspection system recited in claim 6 wherein each of said flat ribs is coated with an X-ray reflective material. 9. The X-ray inspection system recited in claim 6 wherein each of said flat ribs has a plurality of spaced slots for being interleaved with other said flat ribs. 10. A lobster eye lens comprising:a decussate arrangement of interleaved wafer-thin flat ribs configured as a three-dimensional array forming a plurality of contiguous channels of continuously diminishing square cross-section;wherein each of said flat ribs is a segment of an annulus having an inner circumference and an outer circumference. 11. The lobster eye lens of claim 10 wherein said flat ribs are configured to provide said channels of selected length and cross-section for focusing incident X-rays. 12. The lobster eye lens of claim 11 wherein each of said flat ribs has a surface which is made highly reflective to said incident X-rays. 13. The lobster eye lens of claim 10 wherein each of said flat ribs has a plurality of spaced slots for being interleaved with other said flat ribs. 14. A lobster eye lens comprising:a decussate arrangement of interleaved wafer-thin flat ribs configured as a three-dimensional array forming a plurality of contiguous channels of continuously diminishing square cross-section;wherein each of said flat ribs is a segment of an annulus having an inner circumference and an outer circumference and wherein a first plurality of said flat ribs have slots extending from said inner circumference and a second plurality of said flat ribs have slots extending from said outer circumference. 15. The lobster eye lens of claim 14 wherein said decussate arrangement of said interleaved flat ribs is formed by precisely perpendicular mating said first plurality of flat ribs with said second plurality of flat ribs along said slots. 16. The lobster eye lens of claim 15 wherein the thickness of said flat ribs is substantially equal to the width of said slots and the length of said slots is substantially equal to one-half the distance between said inner circumference and said outer circumference along a radius of said annulus. 17. An X-ray imaging system comprising at least one-X-ray source for generating X-ray energy toward a proximate target area, a detector having a focal surface for receiving backscatter X-rays reflected from the target area, and a lobster eye lens for focusing said backscatter X-rays onto said detector focal surface;said lobster eye lens having a decussate arrangement of interleaved wafer thin flat ribs configured as a three-dimensional array forming a plurality of contiguous channels of continuously diminishing square cross-section;said flat ribs being a segment of an annulus having an inner circumference and an outer circumference. 18. An angular staring spectrometer comprising:an X-ray source positioned at a first selected location relative to a sample to be analyzed;a lobster eye lens positioned at a second selected location relative to said sample, said lens having a focusing surface; anda plurality of X-ray detectors located on said focusing surface as an array of detectors selectively spaced from each other to provide an output based upon the cross-sectional geometry of the scattering cone generated by the X-ray backscatter from said sample which is indicative of the material of said sample;said lobster eye lens having a decussate arrangement of interleaved wafer thin flat ribs configured as a three-dimensional array forming a plurality of contiguous channels of continuously diminishing square cross-section;said flat ribs being a segment of an annulus having an inner circumference and an outer circumference. 19. A method of generating an image of objects in an area being inspected; the method comprising the steps of:irradiating said area with X-ray energy;positioning a detector for receiving X-ray backscatter from said irradiated area; andplacing an X-ray lobster eye lens between said area and said detector for focusing said X-ray backscatter onto said detector;said lobster eye lens having a decussate arrangement of interleaved wafer thin flat ribs configured as a three-dimensional array forming a plurality of contiguous channels of continuously diminishing square cross-section;said flat ribs being a segment of an annulus having an inner circumference and an outer circumference. 20. The method recited in claim 19 further comprising the step of configuring said detector for converting said focused X-ray backscatter into an electronic image of said area being inspected. 21. The method recited in claim 20 further comprising the step of locating said lobster eye lens relative to said area being inspected so that the image of said area will be responsive only to objects having elements with low Z-numbers to enhance the contrast of such objects in the image.
052788814	abstract	An Fe-Cr-Mn alloy is disclosed which has the following composition by wt% and corrosion resistance of which is improved and deterioration in its strength is prevented at grain boundaries due to irradiation of high-energy particles such as neutrons: 5 to 40% of Mn, 5 to 18% of Cr, 2.0 to 12% of Al and the balance of Fe except for unavoidable impurities. In the alloy according to the present invention, Al is added to an Fe-Cr-Mn alloy by a restricted quantity as a main component element. As a result of the addition of Al, an alloy can be obtained in which lowering of concentration of Cr at grain boundaries due to irradiation of high-energy particles such as neutrons can be prevented or concentration of the solutes can be raised.
	abstract	A Single Fluid Reactor with an inner zone that includes a solid neutron moderator, which can have through holes defined therein. This solid neutron moderator can have a relatively small diameter, which can range, in some embodiments, from less than one meter to about 1.5 meter. The solid neutron moderator effectively creates an inner zone with a neutron profile that is far more thermalized than if the solid neutron moderator were absent. The surrounding layer of salt surrounding this inner zone has a much harder neutron spectrum.
	claims	1. A scintillator panel comprising:a flexible substrate;a phosphor arranged on the flexible substrate; anda thermal expansion compensation layer disposed between the flexible substrate and the phosphor,whereina linear expansion coefficient of the thermal expansion compensation layer is greater than a thermal expansion coefficient of the phosphor,surfaces, of the thermal expansion compensation layer and of the flexible substrate, in contact with each other each contain an organic substance, andthe thermal expansion compensation layer contains a filler. 2. The scintillator panel according to claim 1, whereina light emitting surface and side surfaces of the phosphor, and side surfaces and a portion of a lower surface of the flexible substrate are covered with a moisture-impermeable protective layer. 3. The scintillator panel according to claim 1, whereina thickness of the thermal expansion compensation layer is greater than or equal to 1 μm and less than or equal to 100 μm. 4. The scintillator panel according to claim 1, whereina phosphor layer is formed from an additive, as a starting material, containing cesium iodide and thallium using a vapor deposition technique. 5. The scintillator panel according to claim 1, whereinthe flexible substrate contains at least one of a resin selected from the group of: polyethylene terephthalate, polyethylene naphthalate, cellulose acetate, polyamide, polyimide polyether-imide, epoxy, polyamide-imide, bismaleimide, fluororesins, acrylic resins, polyurethane, aramide, nylon, polycarbonate, polyphenylene sulfide, polyether sulfone, polysulfone, polyether ether ketone, and liquid crystal polymer, or a carbon fiber reinforced resin; or the flexible substrate is a thin film glass having a surface coated with a resin layer. 6. The scintillator panel according to claim 1, whereinthe thermal expansion compensation layer contains at least one polymer selected from the group of: polyurethane, vinyl chloride copolymers, vinyl chloride-vinyl acetate copolymers, vinyl chloride-vinylidene chloride copolymers, vinyl chloride-acrylonitrile copolymers, butadiene-acrylonitrile copolymers, polyamide resins, polyvinyl butyral, polyester, cellulose derivatives (nitrocellulose), styrene-butadiene copolymers, various synthetic rubber-based resins, phenolic resins, epoxy resins, urea resins, melamine resins, phenoxy resins, silicone resins, acrylic resins, and urea formamide resins. 7. A radiation detector comprising:the scintillator and according to claim 1 as a line sensor. 8. The radiation detector according to claim 7, whereinthe scintillator panel and a plurality of sensors are used in combination.
053316797	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS A first preferred embodiment of a fuel spacer according to the present invention will be described hereunder with reference to FIG. 1 to 6. Referring to FIG. 1, showing a fuel assembly to be incorporated in a boiling water reactor, the fuel assembly 1 comprises an upper tie plate 2 to which a handle 3 is secured, a lower tie plate 4 disposed on the opposite side of the upper tie plate 2, a number of fuel rods 5 which are supported at both axial ends by the upper and lower tie plates 2 and 4, water rods 6 arranged within an arrangement of the fuel rods 5, and a plurality of fuel spacers 9a and 9b arranged in the axial direction of the fuel rods 5. The fuel rods 5 and the water rods 6 are supported by the fuel spacers 9a and 9b with their horizontal distances being kept constant with each other. These fuel rods 5 and the water rods 6 are assembled in a bundle by the fuel spacers 9a and 9b and the bundle is surrounded by a channel box 10, which is secured to the upper tie plate 2. Several fuel rods 5 may be replaced with fuel rods 5a each having a short axial length, called short fuel rod 5a hereinlater, and only the lower ends of the short fuel rods 5a are supported by the lower tie plate 4. The water rods 6 may also be supported only at its lower end by the lower tie plate 4. The channel box 10 maybe secured to the lower tie plate 4 instead of the upper tie plate 2. To the lower end portion of each of the water rods 6 is formed a coolant flow-in port 7 and to the upper end portion thereof is formed a coolant flow-out port 8. As shown in FIGS. 2A', 2B and 3 the fuel spacers 9a and 9b are formed by tubular ferrules 12 of the numbers corresponding to the numbers of the fuel rods 5 and the short fuel rods 5a in a lattice arrangement and the outer periphery of the bundle of the tubular ferrules 12 is surrounded by a support band 11 in the shape of a belt. The lattice arrangement of the tubular ferrules 12 are joined with adjoining lattice arrangement thereof by means of spot welding. One end of the tubular ferrule 12 is formed so as to provide four trapezoidal or triangular petal portions with cutting off some part of end portion, separated by equal distances in the circumferential direction of the tubular ferrule 12. In the illustration of FIG. 4, the trapezoidal petal portions 21 and cutouts 20 are formed around the end portion by cutting out some parts of the end portion of the ferrule 12. Two projections 13a projecting inward from the ferrule 12 are formed thereto at portions below the trapezoidal petal portions 21 and other projections 13b are also formed to the ferrule 12 at portions near the other end portion to which any cutout portion is not formed. These projections 13a and 13b are formed by inwardly projecting portions of the tubular wall of the ferrule itself. The ferrules 12 each of the structure shown in FIG. 3 are arranged in a lattice in alternatingly reverse, i.e. upside down, axial attitudes. As shown in FIG. 4, each ferrule 12 provides four petal portions 21 each having trapezoidal shape and portions between the adjoining petal portions 21 are formed each having a flat bottom portion 22 having a point 23 at which the ferrule 12 is joined by means of spot welding with the point 23 of an end portion provided with no petal portions of an adjacent another ferrule 12. The points 23 of the other ends of the ferrule 12 as shown in FIG. 4 opposite the end having the petal portions 21 are also joined by means of spot welding with points 23 of the flat portions 22 of another ferrule 12 disposed adjacently. This state is shown in FIG. 5, from which the lattice arrangement of the ferrules 12 will be understood with the assistance of FIG. 3. As shown in FIGS. 3 and 5, the number of ferrules 12 are arranged in a lattice with adjoining ferrules 12 are reversely arranged in their axial attitudes and respective adjoining ferrules 12 are joined with each other by means of spot welding at points 23 formed to both axial ends of the respective tubular ferrules 12. FIGS. 6A and 6B are an elevational view and cross sectional view, respectively, showing a relationship between the fuel rods 5 and 5a and the tubular ferrules 12. Referring to FIGS. 6A and 6B, a continuous loop spring 14 is disposed so as to span the adjoining two tubular ferrules 12 and the loop spring 14 has an outwardly projected portion at its vertically central portion as shown in FIG. 6A. The continuous loop spring 14 is attached to the ferrule 12 at its one end so that the one end is engaged with the pawl piece 16 projecting inside the cutout portion 15 formed substantially rectangularly to the side wall of the tubular ferrule 12 as shown in FIG. 4. The pawl piece 16 is formed to a portion near one end, i.e. lower end as viewed, of the rectangular cutout portion 15. The other one end of the loop spring 14 is attached to the pawl piece 16 of the adjoining tubular ferrule 12 arranged in an axially upside down attitude, and so the loop spring 16 for pressing the fuel rods in the fuel assembly are supported between the adjoining two tubular ferrules 12 arranged in axially reversed attitudes. Accordingly, it is to be noted that the respective tubular ferrules 12 are arranged in a lattice in their vertical attitudes in which one ferrule 12 has the upper end provided with the petal portions 21 and another one adjoining to this one ferrule 12 has the lower end provided with the petal portions 21. According to this embodiment, as shown in FIG. 5, the structure of each of the fuel spacers 9a and 9b has a central vacant portion 18 formed by removing seven ferrules 12 arranged at the central portion of the fuel spacer 9a(9b). Namely, at the central portion of the fuel spacer is formed an elongated vacant portion 18 which is defined by side walls of centrally arranged ten tubular ferrules 12. Two bridging members or pieces 17a and two bridging members or pieces 17b are respectively disposed diagonally opposingly inside the elongated vacant portion 18. Both the ends of the bridging pieces 17a are secured, by spot welding means, to the side walls of the adjoining two ferrules 12, constituting both ends of the elongated vacant portion, of the ten ferrules 12 defining the vacant portion 18. Each of the bridging pieces 17a is composed of upper and lower two plate members having a recessed portion formed to an upper or lower end to which the continuous loop spring 19 is mounted. This portion is welded after the assembling thereby to form one bridging piece 17a, and thereafter, the bilateral end portions of the bridging piece 17a are spot welded to the upper and lower ends of the tubular ferrule 12. Each of the other bridging pieces 17b combines three tubular ferrules 12 constituting the above-mentioned ten ferrules 12 and forms a bent portion 31 projecting in a diagonal direction of the fuel spacer 9a (9b), and the water rods 6 are supported by the corner portions of this bent projected portion 31. Namely, in the illustration of FIG. 5, two water rods 6 are supported by the corner portions of the projected portions 31 of the two bridging pieces 17b disposed diagonally in the fuel spacer. The bilateral ends and the root of the bent projected portion of each of the bridging pieces 17b are formed so as to have substantially the same curvature as that of the side wall of the tubular ferrule 12, and at these portions are spot-welded the upper and lower end of the bridging piece 17b to the adjoining ferrules 12. As shown in FIG. 2 and FIG. 5, two water rods 6 are arranged in the central vacant portion 18 and press supported to the bent portions 31 of the bridging pieces 17b by means of continuous loop springs 19 provided for the bridging pieces 17a. The fuel spacer 9b shown in FIG. 2A is to be attached to the upper portion of the fuel assembly 1, in which the short fuel rods 5a is formed as a vacant portion. In the arrangement of the tubular ferrules 12 described above, as shown in FIG. 3, to the outer periphery of the bundled ferrules 12 are secured a periphery support band member 11 to which a plurality of projections 32 directed upward are formed, the projections 32 being at least inwardly bent at an intermediate portion of adjoining fuel rods 5. A pair of lobes 33 are also formed integrally with the support band member 11 at portions near the four corners thereof and so as to project outwardly to keep a distance constant between the support band member 11 and the channel box 10 surrounding the same. As described above, according to the first embodiment, it is characterized that the petal portions 21 formed to one end of the tubular ferrules 12 are projected upwardly or downwardly from the flat portions of the upper and lower edges of the periphery support band member 11, but in a modification, the petal portions 21 may not project over the upper and lower edges of the support band member 11 by making the height of the support band member 11 higher. The first embodiment of the present invention operates as follows. Reactor cooling water, called herein a coolant, is fed from the lower side of a reactor core into a fuel assembly charged in the core and raises therein. That is, the coolant is first flown into the lower tie plate 4 from the lower side thereof and then rises in a flow passage formed between the fuel rods 5 and 5a in the channel box 10. Finally, the coolant flows out into an upper plenum through the upper tie plate 2. A part of the coolant flown inside the channel box 10 is fed into the water rod 6 through the coolant flow-in port 7 formed thereto and then raises upward and flows outward through the coolant flow-out port 8 also formed to the water rod 6. As described, the coolant flown in the water rod 6 increases the percentage of the coolant at the central portion in cross section of the fuel assembly 1, thereby improving the moderation of neutrons at its central portion. For this reason, the reactivity at the central portion in the cross section of the fuel assembly 1 is made high and the power distribution in the cross sectional area is also made flat. The coolant removes the heat of the fuel rods 5 while rising axially upward in the channel box 10 and, hence, the coolant itself is heated to a temperature from a sub-cooled state to a saturated state, and causes the boiling of the saturated water. Accordingly, ideally, the most desirable heat removing function will be achieved in a state in which the liquid phase (saturated water) in the coolant flow near the surface of the fuel rods and the steam phase flow in a space between the fuel rods 5 and 5. On the other hand, in view of the actual flow condition of the coolant in the fuel assembly, steam-liquid two phase flow called as circular flow is realized in the upper half area of the fuel assembly in which a void fraction is high and there is less margin in the heat removal. In such flow state, as shown in FIG. 7, the void fraction is made high at a space between the fuel rods 5 and 5a and the surfaces of the fuel rods 5 are covered with liquid films 27. The heat can be removed by the boiling of these liquid films 27. In FIG. 7, reference numeral 28 indicates a liquid drop and numeral 29 denotes steam voids. Then, referring to Figs, 8A, 8B and 8C, in the fuel spacer of the prior art in which lower ends of the side walls of the tubular ferrules are all in the same level, when two-phase flow collides with the fuel spacers, horizontal vector of the two-phase flow in the fuel rod cell at the lower ends of the fuel spacers becomes large and the liquid films on the surfaces of the fuel rods near the lower ends of the fuel spacers are largely constricted thereby and make worse the heat removing function, whereat the transition boiling may likely be caused. The variation of the projected area of the fuel spacer becomes rectangular as shown in FIG. 8A, whereby the variation of the rapid-constriction and rapid-expansion is made large and the local pressure loss is also made large. On the contrary, according to the structures of the fuel spacers of the present invention, as shown in Figs. 9A, 9B and 9C, the cross sectional area of the lower end of the fuel spacer is substantially trapezoidal and can be made less than one sixth (about 1/6 in this embodiment), so that the horizontal velocity vector of the steam-liquid two phase at a portion near the lower end of the fuel spacer becomes small in comparison with the conventional fuel spacer and the constriction of the liquid film 27 on the surface of the fuel rod 5 can be made small. Although the flow cross sectional area reduces gradually during the passing through the cutout portions of the end portions of the ferrules, during this passing, the flow in the tubular ferrule is guided into a flow outside the tubular ferrule having a small resistance thereby to increase the flow velocity in this flow passage 25. Consequently, the flow velocity at the space portion 24 between the fuel rods 5 in the tubular ferrule becomes small in comparison with the case of the prior art of FIG. 8 and the liquid film is made thicker. Thereafter, with reference to FIG. 9A, the steam-liquid flow reaches the full-thickened portion Y of the fuel spacer, whereat the change of the thickness thereof is about 2/3 in comparison with the conventional one and the flow distribution is made in the area X, and accordingly, the disturbance of the flow at this area Y is made small in comparison with the prior art. As this result, the constriction of the liquid film 27 can be made small. When the flow reaches the area Z, the cross sectional area of the fuel spacer is constricted to about 1/3 from the fully thickened state, and thereafter, the cross sectional area thereof is gradually reduced, so that the pressure loss due to the rapid expansion of the cross sectional area of the fuel spacer can be reduced. Furthermore, the axial height of the fuel spacer is made large in appearance in comparison with the conventional one, and the frictional pressure loss increases, but the increasing of the frictional pressure loss of such extent in appearance does not give any significant adverse affect, and on the contrary, the pressure loss due to the constriction or expansion of the flow passage may be more adversely effected. Moreover, in the described embodiment, the portions 23 to be spot welded are formed to portions on the intermediate flat portions 22 between the respective petal portions 21 formed to the end of the tubular ferrule 12, thus the spot-welding being easily carried out, and accordingly, in spite of the specific structure of the present invention, the welding assembling workings of the fuel spacer are not made complicated in comparison with the conventional workings. Consequently, according to this first embodiment, the shape of the cross sectional area of the tubular ferrule with respect to the flow direction of the steam-liquid phase provides a smooth trapezoidal shape but not rectangular shape as in the conventional ferrule. Therefore, the local pressure loss due to the rapid constriction or rapid expansion of the flow passage at the fuel spacer can be made small, and moreover, the horizontal turbulent flow near the surface of the fuel rod 5 supported by the fuel spacer can be also made small in comparison with the conventional structure of the fuel spacer, so that the liquid film 27 on the surface of the fuel rod 5 is not peeled off or has less constriction. As this result, the pressure loss can be made small in comparison with the fuel spacer utilizing the tubular ferrules, thus contributing the improvement of the critical power output of the fuel assembly. A second embodiment of the fuel spacer according to the present invention will be described hereunder with reference to FIG. 10. Although in the first embodiment, the fuel spacer has one end, lower end as viewed, being made flat, in this second embodiment, the ferrule 12b constructing the fuel spacer has the lower, as viewed in FIG. 10, end provided with triangular cutout portions 26 with projected portions and spot-welded portions being disposed between the adjacent triangular cutout portions 26. According to this embodiment, the cross sectional shape of the fuel spacer can be made further smooth more than that of the first embodiment. In this second embodiment, by utilizing the tubular ferrules 12b of such structure as shown in FIG. 10, the axial cross sectional area of the fuel spacer can be made more smooth and the local pressure loss can be hence further reduced. However, since the location of such cutout portions 26 reduces the strength of the tubular ferrule 12b, it is not desired to form too large cutout portions in size and number. In view of this fact, according to the second embodiment shown, two triangular cutout portions 26 are formed at portions other than the welding portions 23 and the projected portions 13b facing the flow passage 25 surrounded by the outer wall of four tubular ferrules in the square lattice arrangement of the tubular ferrules 12b. A third embodiment according to the present invention will be next described hereunder with reference to FIG. 11, showing a perspective view of a tubular ferrule 12c for the fuel spacer. In this third embodiment, V-shaped or M-shaped petal portions 21a are further formed to the trapezoidal petal portions 21 of the first embodiment. In the illustration, the lower end of the tubular ferrule 12c is made flat, but triangular cutout portions such as those 26 in the second embodiment may be additionally formed. According to the structure of the fuel spacer utilizing the tubular ferrules 12c each shown in FIG. 11, the flow colliding with the petal portions 21a at the upstream side of the fuel spacer (in the illustration, the upper end) flows as shown by arrows in a circumferential direction along the outside and inside of the tubular ferrule 12c and along the shape of the petal portions 21a. In this embodiment, the petal portions 21a have a shape of small flow resistance and ready for guiding the steam-liquid flows into the flow passage 25 surrounded by the outer wall of the tubular ferrules 12c. On the contrary, in the embodiment of FIG. 4, the trapezoidal petal portions 21 have shaped ready for guiding to the portions 23 to be spot-welded to adjoining tubular ferrule 12, and since this portion constitutes a narrow flow passage, the flow resistance increases. Accordingly, the fuel spacer of the third embodiment is more effective in this meaning than that of the first embodiment. FIG. 12 shows a fourth embodiment of the fuel spacer composed of tubular ferrules according to the present invention. In this fourth embodiment, two kinds of ferrules are combined and, namely, in the illustration of FIG. 12, the ferrule 12a of righthand side has an upper portions 21b and a lower end having flat portion to which (downstream) end formed to have trapezoidal petal portions 21b and a lower end having flat portion to which a welding point 23 is formed. The ferrule 12c of lefthand side has an lower (upstream) end formed to have V- or M-shaped petal portions 21a as shown in FIG. 11 and an upper end having a flat portion to which a welding portion 23 is formed. These ferrules 12a and 12c are welded together at their welding points 23. The petal portions 21b of the righthand ferrule 12a are twisted in one direction outward of the ferrule 12a to form an opened upper end. According to the combined arrangement of the ferrules 12a and 12c of FIG. 12, the V- or M-shaped petal portions of the lower end, upperstream side end, of the ferrule 12c give less local pressure loss, and the trapezoidal shape of the petal portions 21b at the downstream side of the ferrule 12a is suitable for imparting a swirlling force to the two-phase flow in the flow passage surrounded by the side wall of the ferrule 12a. Liquid drops of the two-phase flow in this passage deposite to the surfaces of the fuel rods arranged near by centrifugal force caused by the swirlling force. The liquid drops in this flow passage flow at fast speed in the passage between the fuel rods after the passing through the fuel spacer area, so that the liquid drops contained in this flow raise in that flow passage without contributing to the heat removal of the fuel rods, thus achieving adverse cooling effect. In this embodiment, the petal portions 21b of the righthand ferrule 12a are twisted and opened outward by imparting the revolutional force thereby to enhance the cooling effect, making delay the generation of the transition boiling at the downstream side of the fuel spacer and hence contributing to the improvement of the critical power output. FIG. 13 represents a fifth embodiment according to the present invention, in which a tubular ferrule 12d is provided with both ends to which petal portions are formed in spite of the fact that the petal portions are formed to only the one end of the ferrule in the foregoing embodiments. Namely, referring to FIG. 13, petal portions 21a and 21b are formed to both the ends of the tubular ferrule 12d, and in the illustration, for example, the upstream (lower) side end of the ferrule 12d has V- or M-shaped petal portions 21a and the downstream (upper) side end thereof has trapezoidal petal portions 21b as shown in FIG. 12 for achieving substantially the same effect as that of the fourth embodiment of FIG. 12. According to this fifth embodiment, the change of the cross sectional area of the fuel spacer can be made most smooth and the pressure loss can be made small. The outwardly twisted structure of trapezoidal petal portions formed to the downstream side end of the tubular ferrule for the individual fuel rod cell further acts as a guide for the smooth insertion of the fuel rod at a working time for charging the fuel rods into the fuel spacer in an assembling of the fuel assembly. Although, in the foregoing embodiments, there are proposed tubular ferrules of substantially circular shape as basic structure, octagonal tubular ferrules such as shown in FIG. 16 may be utilized as the basic structure thereof. Furthermore, according to the fuel spacer composed of the tubular ferrules of the present invention, the fuel assembly including fuel rods in 9.times.9 lattice arrangement including two water rods are described, but the present invention can be applied to another lattice type arrangement such as 8.times.8 fuel rod arrangement with various shapes or numbers of water rods. It is also to be noted that the present invention is not limited to the described embodiments or examples, and many other changes or modifications may be made within the scope of the appended claims.
	abstract	To provide an X-ray microscopic inspection apparatus capable of performing non-destructive inspection with high resolving power within a very short period, and having advantageous functions such as a high precision electron probe control function, a CT function, an elemental analysis function, and a target switching function. The apparatus includes a magnetic superposition lens having a magnetic field generating portion disposed in the vicinity of an electron generating portion of an electron gun; reflected electron detecting means having a detecting portion disposed above the target for X-ray generation, for detecting a reflected electron from the target; and electron image generating means for performing imaging of a target surface utilizing the signals from the reflected electron detecting means, wherein the apparatus is arranged so that alignment including focus adjustment to the target for X-ray generation and astigmatism correction may be performed based on image information of the electron image. Further, the apparatus is equipped with functions such as the electron probe control function, the CT function, the electron axis alignment function the elemental analysis function, and the target switching function.
	description	This invention relates generally to computer system performance studies, and more particularly to the collecting and analyzing of computer system performance data. Nowadays, a computer system has evolved into a complicated combination of multiple software and hardware components for performing various functions and supporting various features. To obtain optimal performance of a computer system, continuous monitoring of the performance of the computer system and/or its components is necessary. Continuous studies of component performance are necessary, not only in operating an existing computer system, but also in developing computer software and hardware components. For example, when developing an operating system, such as the Microsoft® Windows® XP operating system, development teams of various components of the operating system constantly stress test the various components. Stress testing is the process of subjecting a component to strenuous operating conditions and observing whether the component can withstand heavy usage without failure. Stress testing thus helps a component development team to identify any weakness or defect in the component and can provide valuable information as to the causes of a failure if meaningful data are collected during the stress testing process. Therefore, an effective study of system performance should be able to collect meaningful data regarding the operating characteristics of a component and make the data easily accessible. During stress testing or other performance studies, a component may have a number of statistical variables that are of interest and should be tracked. For a computer system component, such statistical variables can capture the usage and information with regard to system memory, CPU, event log, etc., in the component. The collected data can then be used to identify the status of the component and diagnose problems in the component. Conventionally, such statistical data concerning a component in a computer system is called a metric. Conventional approaches in capturing metrics usually provide static snapshots of the current status of a component performing a task such as stress testing. However, a static snapshot fails to reflect changes of a metric over time. Moreover, conventional approaches usually collect metrics that are specific for an individual computer system or component, rather than metrics that are common to different computer systems or components. Thus, conventional approaches fail to reveal how the same metric may vary in different computing environments. Further, collected metrics are usually stored as a text report, which provides a user little flexibility or variation in presenting metric data. Furthermore, different component teams may store collected metrics in different formats and in different locations, rather than in a uniform format and at a centralized location that everyone can access and use. As a result, metric data provided by one component team often cannot be easily integrated with metric data provided by other component teams, therefore making it difficult to establish consistent system evaluation of different components. Therefore, there exists a need to collect any system metric during run time of a component in a predefined format, and to store the system metric in a way so as to enable analyzing of either the individual metric or a combination of metrics, either offline or online. This invention addresses the above-identified need by providing a framework to collect any system metrics during run time in a predefined format and to store the system metrics in a way so as to enable the analysis of either an individual metric or a combination of metrics, online or offline. The invention enables development teams to record vital systems and any other performance metrics at any given interval and to analyze trends in each of the metrics with regard to time. One aspect of the invention provides multiple computer-executable modules: a configuration module, a data collection module, a central metrics storage module, and preferably an analysis module. A configuration module configures one or more metrics. The configuration module includes a configuration file and preferably a user interface through which a user may configure one or more metrics (hereinafter “configuration UI”). The configuration file specifies the name of a metric and a log file on a client system for recording metric data during run time of the client system. The configuration file also specifies a time interval for periodically collecting metric data and uploading them to the central metrics storage module, which can be a central database for storing all metric data. The configuration file further specifies the type of information contained in a metric. A metric may include events occurring during the run time of a client system. A metric may also include performance counters or data on drivers on the client system. A metric may further include one or more processes running on the client system. According to settings in the configuration model, the data collection module collects one or more metrics concerning one or more computer components on one or more client systems. The data collection module sends the collected metric data to the central metrics storage module that centrally stores all metric data. The central metrics storage module may format the received metric data prior to storing them. Preferably, the invention also provides an analysis module that analyzes the stored metric data online or offline. The analysis module may analyze the metric data statistically or chart one or more metrics graphically to show the pattern over time. The analysis module may further include a user interface (hereinafter “analysis UI”), through which a user analyzes and/or graphs stored metric data. Another aspect of the invention extends and customizes the functionalities provided by the above-mentioned modules with one or more component specific plug-ins. A component specific plug-in can be a configuration UI plug-in, a data collection plug-in, or an Intelligent Pass/Fail plug-in (hereinafter “IPF plug-in”). A configuration UI plug-in extends the configuration UI with component specific configuration information. A data collection plug-in extends and customizes the data collection module. An IPF plug-in includes both an IPF client plug-in and an IPF UI plug-in. The IPF client plug-in receives collected metric data from the data collection module and decides whether the metric data meet one or more criteria in the IPF client plug-in. If the collected metric data do not meet the criteria in the IPF client plug-in, the IPF client plug-in returns a value indicating a failure; the data collection module then discontinues collecting the metric and exits the task that generates the metric data collected (hereinafter “metric task”). If the metric data meet the criteria in the IPF client plug-in, the IPF client plug-in returns a value indicating a pass; the data collection module then continues collecting the metric. In summary, the invention provides a framework that collects, stores, and analyzes metric data concerning a client system or component. Component development teams can use the framework to monitor and improve component performance, scalability, and reliability. IT professionals can use the framework to build critical system monitoring tools for monitoring the health of infrastructure servers. Embodiments of the invention provide a framework for collecting metrics during run time of a client system or component in a predefined format and for storing the metrics in a way so as to be able to analyze either an individual metric or a combination of metrics online or offline. The invention thus records vital system and other performance metrics at given intervals and analyzes trends in the metrics with regard to time. In general, a metric is an abstract term referring to any one of system performance counters, system resource usage information, etc. A metric usually consists of a name and an associated value. For example, in a Microsoft Windows® operating system, one of the system metrics is the usage of memory by various system processes such as winlogon.exe, service.exe, etc. An exemplary embodiment of the invention defines the winlogon.exe process with the “metric name” as “winlogon.exe/paged pool bytes” and “metric value” as the actual paged pool memory bytes used by the winlogon.exe process. Exemplary embodiments of the invention contain multiple phases: defining a metric, collecting and/or monitoring the metric, storing the metric, and/or analyzing the metric. FIG. 1 is a block diagram illustrating these multiple phases. In a “defining metric” phase 102, the invention defines any system run time information as a metric name/value pair and configures the settings for each metric. In a “collecting/monitoring metric” phase 104, the invention collects the defined metrics during run time of a client system or component at defined time intervals. Preferably, the invention also monitors the collected metric data. The invention may perform an independent action, such as stopping executing the relevant metric task when an individual metric does not meet one or more specified criteria. The one or more criteria for taking certain actions can be applicable to an individual metric or a combination of metrics. In a “storing metric” phase 106, the invention stores collected metrics in a central database in a defined format. The invention thus enables proper archiving of all metrics for any future use, such as analyzing them to find trends in system behavior over time. Preferably, the invention further includes an “analyzing metric” phase 108, during which the invention retrieves the stored metrics for both online and offline analysis. Analysis of metrics may include charting trends of individual or collective metrics on individual or collective client systems. These charts can then be used to improve scalability, performance, and reliability of individual components by different component development teams. Embodiments of the invention provide an architectural framework for implementing the functionalities provided by the multiple phases of the invention, such as the phases illustrated in FIG. 1. FIG. 2 is a block diagram illustrating an exemplary framework 200. Different modules in the framework 200 perform the functionalities provided by the multiple phases of the invention. The framework 200 includes a configuration module 202, a data collection module 208, a central metrics storage module 218, and, preferably, an analysis module 220. In addition, the framework 200 may further include one or more component specific plug-ins 210 that extend and customize the framework 200. As illustrated in FIG. 2, the data collection module 208 reads setting information from the configuration module 202, and preferably interacts with one or more component specific plug-ins 210, to collect one or more metrics on one or more client systems 212. In some embodiments of the invention, the data collection module 208 interacts with an execution agent 214 on a client system 212 to obtain information on the job that a metric task runs under. The execution agent 214 links the metric with the job running on the client system 212. The data collection module 208 then sends, via a communication layer 216, collected metric data to a central metrics storage 218. If an analysis module 220 exists, the analysis module 220 can analyze the metrics after retrieving one or more metrics from the central metrics storage 218 at any time, i.e., during or outside the run time of the client systems 212. Specifically, the configuration module 202 configures settings for one or more metrics. In embodiments of the invention, the configuration module 202 includes a configuration file 204 and preferably a configuration UI 206. The configuration file 204 contains the settings for each individual metric. For example, the configuration file 204 defines the name of a metric, the log file for storing collected metrics on a client system, and the time interval of updating the collected metrics to the data storage module 218. The configuration file 204 further specifies what a metric contains. In embodiments of the invention, a metric can be different types of events generated by a client system or component. A metric can also be a performance counter that records status information about various system devices or processes. A metric can further be any of the available driver data such as PoolTags on the client system that provide information on how drivers use available system resources. A metric can also be any of the available processes running on a client system. In addition, the configuration file 204 also identifies one or more component specific plug-ins 210 associated with a metric. As noted above, a component specific plug-in 210 extends and customizes the framework 200 in collecting one or more metrics. FIG. 5 illustrates exemplary component specific plug-ins 210 and will be discussed in detail later. As noted above, according to the settings in the configuration file 204, the data collection module 208 collects a metric and updates the central metrics storage module 218 periodically. In some embodiments of the invention, the configuration module 202 further includes a configuration UI 206 that enables a user to configure one or more metrics via the user interface. The configuration UI 206 outlines the setting requirements of the configuration file 204. A user's input to the configuration UI 206 is deposited into the configuration file 204. FIGS. 3A–3B illustrate one exemplary configuration UI 206. The configuration UI 206 contains individual sections logically grouping the various settings concerning a metric. These individual sections appear as individual tabbed pages in the configuration UI 206. Specifically, the configuration UI 206 contains a general section 302 (FIG. 3A). The general section 302 identifies the name 304 of an output log file on a client system 212. The output log file 304 stores collected metrics on the client system 212. The general section 302 allows a user to specify a report interval 308 for collecting metric data on the client system 212, storing the metric data in the output log file 304, and reporting the metric data to a central metrics storage module 218 (FIG. 2). The general section 302 may further provide a “Do Not Report to DB” check box 310. The actuation of the check box 310 disables the periodical sending of metric updates to the central metrics storage module 218; the collected metric data remains locally in the output log file 304 on the client system 212. The configuration UI 206 also contains sections for specifying what information one or more metrics may include. As noted above, a metric may be any of the events occurring during the run time of a client system 212, performance counters, driver PoolTags, and processes on the client system 212. The configuration UI 206 includes an event log section 314 (FIG. 3B) to set configurations for copying events from the event logs of a client system 212 and forwarding them to the central metrics storage module 218. The event log section 314 enables a user to choose to monitor specific types of events that occur on a client system 212. As shown in FIG. 3B, the event log section 314 contains a list of check boxes, each of which represents an option to collect a specific type of event log entry. For example, the types of event log entries can be error 322, audit success 324, audit failure 326, information 328, and warning 330. The event log section 314 also includes a “report all events” check box 332 that allows for reporting all existing events in an event log, regardless of its event type. The configuration UI 206 further includes a PerfMon section 334 (FIG. 3A, details not shown) for setting configurations to collect performance-related data exposed through one or more performance counters registered on a client system 306. In an exemplary embodiment of the invention, the PerfMon section 334 lists all registered performance counters from which a user may select one or all of them to collect metric data on. The configuration UI 206 also contains a PoolTags section 336 (FIG. 3A, details not shown) that sets configurations to collect memory usage information for various drivers in the client system 212 by using their corresponding PoolTags. As known by those of ordinary skill in the art and other related fields, a driver PoolTag identifies how a driver in a client system uses available computer resources such as computer memory. In an exemplary embodiment of the invention, the PoolTags section 336 lists all available PoolTags along with their descriptions; a user may select to exclude or include specific PoolTags for data collection. The configuration UI 206 further contains a process section 338 (FIG. 3A, details not shown) that allows a user to set configurations for collecting information about various processes in the client system 212. In an exemplary embodiment of the invention, the process section 338 identifies available processes that a user may choose to collect metric data on. In some embodiments of the invention, the configuration UI 206 may be extended by a configuration user interface provided by an IPF UI plug-in (hereinafter “IPF configuration UI”). The IPF configuration UI allows a user to specify thresholds for one or more metrics so as to determine whether the corresponding metric task has succeeded or failed according to the thresholds at run time of one or more client systems 212. FIG. 4 illustrates one exemplary IPF configuration UI 340. The IPF configuration UI 340 contains a preview panel 342 displaying all available metrics that have been selected in the various configuration UI 206 sections previously discussed. For example, as shown in FIG. 4, a “pool paged allocation” metric 344 is selected in the PoolTags section 336 (FIG. 3A). The IPF configuration UI 340 further contains a “Pass/Fail Criteria” panel 345. The “Pass/Fail Criteria” panel 345 includes a “pass” radio button 346, a “fail” radio button 348, and a criterion list view 350 that lists one or more criteria used to judge whether a metric task that the selected metric concerns fails or succeeds. A criterion defines a threshold for a given metric. If the “pass” radio button 346 is selected, the criteria in the criterion list view 350 are used to judge whether the metric task has succeeded. If the “fail” radio button 348 is selected, the criteria in the criterion list view 350 are used to judge whether the metric task has failed. In an exemplary implementation of the criterion list view 350, each criterion entry in the criteria list view 350 identifies the name 354 of the selected metric. The entry also identifies an operator 356 that is used for measuring the selected metric against a threshold 358 at run time. The operator 356 can be “>”, “≧”, “<”, “≦”, “=”, and “!=”, etc. The entry specifies the threshold 358 for judging whether the metric task has succeeded or failed. In an exemplary embodiment of the invention, the threshold 358 is a numerical value. For example, in FIG. 4, the threshold value 358 is set to be 33333. The entry may further include a logical operator 360 that is used to join multiple criteria. The logical operator 360 can either be an “AND” or an “OR.” As noted above, the data collection module 208 collects and reports all the required metric data from one or more client systems 212 according to the settings specified in the configuration module 202, preferably along with one or more component specific schema defined in one or more component specific plug-ins 210. Preferably, the data collection module 208 formats all the collected metric data appropriately, for example, into an XML format. The collected metric data is then packed into a database message and is transported via a communication layer 216 to the central metrics storage module 218. More specifically, the data collection module 208 collects metrics based on the settings in the configuration file 204. For example, the data collection module 208 collects all events according to the one or more event types specified in the configuration file 204. For instance, if the configuration file 204 specifies to collect metrics on events of types such as error 322 and audit success 324, the data collection module 208 collects all events with types matching “error” and “audit success.” Unless the configuration file 204 specifies to “report all events” 332 (FIG. 3B), the data collection module 208 will only report new events that have been generated after the data collection module 208 is activated. In an exemplary embodiment of the invention, the data collection module 208 includes a specific identification (number or otherwise) within the metric data sent to the central metrics storage module 218. The identification may include items such as machine name or other hardware/software based identification mechanism that will allow the analysis component 220 to access data specific to a machine or a collection of machines from the central metrics storage module 218. As mentioned before, exemplary embodiments of the invention support one or more component specific plug-ins 210. A component specific plug-in 210 extends and customizes the framework 200. It provides a set of interfaces that different components development teams can implement to extend and customize the framework 200 to support a component's specific characteristics. Embodiments of the invention support at least three types of component specific plug-ins. FIG. 5 is a block diagram illustrating the three types of component specific plug-ins 210 supported by an exemplary embodiment of the invention. The first is a configuration UI plug-in 502. The configuration UI plug-in 502 extends and customizes the configuration UI 206 to create, add, and update custom configuration data concerning one or more metrics for a component. The second type of component specific plug-ins 210 is a data collection plug-in 504 that extends and customizes the data collection module 208. The data collection plug-in 504 receives configuration information from the data collection module 208 and collects custom metric data that is pre-defined in the data collection plug-in 504. The data collection plug-in 504 then exports the collected metric data to the data collection module 208 for it to transport to the central metrics storage module 218. The third type of component specific plug-ins 210 is an IPF plug-in. Embodiments of the invention provide two types of IPF plug-in: an IPF client plug-in 506 and an IPF UI plug-in 508. An IPF client plug-in 506 is loaded into the data collection module 208 during run time and monitors the actual metrics collected at run time to ensure they are within one or more defined criteria. The IPF UI plug-in 508 is loaded into the configuration UI 206 to enable a user to customize the settings for the IPF client plug-in 506. The above-discussed FIG. 4 provides one exemplary user interface provided by an IPF UI plug-in 508. If a metric is associated with at least one IPF client plug-in 506, the data collection module 208 will pass the current metric data it collects to the IPF client plug-in 506 before the data collection module 208 sends the updated metric data to the central metrics storage module 218. The IPF plug-in 506 decides whether the collected metric data meet one or more criteria, for example, specified by a user through the configuration UI 206 (FIG. 4). If the IPF client plug-in 506 returns a value indicating a pass, the data collection module 208 may log the pass result in the appropriate log file for the metric task and continue the metric task. If the IPF client plug-in 506 returns a value indicating a failure, the data collection module 208 logs the failure result in the appropriate log file for the metric task, exits the metric task, and deletes the metric data without reporting them to the central metrics storage module 218. Embodiments of the invention allow more than one IPF client plug-in 506 to interact with the data collection module 208. A failure result from any one of the IPF client plug-ins 506 will cause the data collection module 208 to exit the metric task after appropriate logging. The configuration file 204 specifies the order in which each of the multiple IPF client plug-ins 506 analyzes the metric at run time. Returning to FIG. 2, after collecting metric data according to the specifications in the configuration file 204 and in one or more of the component specific plug-ins 210, the data collection module 208 sends the metric data via a communication layer 216 to the central metrics storage module 218. In some embodiments of the invention, to allow for more effective and extensible format for storing metric data, metric data is transformed before being stored in the central metrics storage module 218. The transformation is transparent both to the client system 212 and the data collection module 208. In an exemplary embodiment of the invention, the output log file 304 containing metric data in a client system 212 includes a parameter called “group name.” The central metrics storage module 218, which can be a database, includes an attribute called “metric name.” The transformation process prefixes the “group name” to the “metric name” and the resulting string is stored as the actual “metric name.” The following XML text illustrates how metric data provided by the data collection module 208 may be transformed. The process information 338 (FIG. 3A) as seen in the output log file 304 (FIG. 3A) storing metric data is as follows: < Group ⁢ ⁢ ⁢ Name = ′′ ⁢ MEMORYINFO ⁢ \ ⁢ PROCESS ′′ > < Metric ⁢ ⁢ Name = ′′ ⁢ services . exe ⁢ \ ⁢ ProcessID ′′ ⁢ Type = System ′′ ′′ > 444 < / Metric > < Metric ⁢ ⁢ Name = ′′ ⁢ services . exe ⁢ \ ⁢ HandleCount ′′ ⁢ Type = System ′′ ′′ > 0 < / Metric > < Metric ⁢ ⁢ Name = ′′ ⁢ services . exe ⁢ \ ⁢ WorkingSet ′′ ⁢ Type = System ′′ ′′ > 17907712 < / Metric > < Metric ⁢ ⁢ Name = ′′ ⁢ services . exe ⁢ \ ⁢ VirtualBytes ′′ ⁢ Type = System ′′ ′′ > 72728576 < / Metric > < Metric ⁢ ⁢ Name = ′′ ⁢ services . exe ⁢ \ ⁢ PagedPoolBytes ′′ ⁢ Type = System ′′ ′′ > 65088 < / Metric > < Metric ⁢ ⁢ Name = ′′ ⁢ services . exe ⁢ \ ⁢ NonPagedPoolBytes ′′ ⁢ Type = System ′′ ′′ > 13880 < / Metric > ⁢ ⋯ ⁢ < / Group > The same information stored in the central metrics storage module 208 after going through the above-mentioned transformation is as follows: <Metric Name=“MEMORYINFO\PROCESS\services.exe\ProcessID” Type=“System”>444</Metric> <Metric Name=“MEMORYINFO\PROCESS\services.exe\HandleCount” Type=“System”>0</Metric> <Metric Name=“MEMORYINFO\PROCESS\services.exe\WorkingSet” Type=“System”>17907712</Metric> <Metric Name=“MEMORYINFO\PROCESS\services.exe\VirtualBytes” Type=“System”>72728576</Metric> <Metric Name=“MEMORYINFO\PROCESS\services.exe\PagedPoolBytes” Type=“System”>65088</Metric> <Metric Name=“MEMORYINFO\PROCESS\services.exe\NonPagedPoolBytes” Type=“System”>13880</Metric> After storing various metrics collected at run time in the central metrics storage module 218 in a defined format, some embodiments of the invention provide an analysis component 220. The analysis component 220 can be used to query the central metrics storage module 218 and to analyze the retrieved metric data, both online or offline. The analysis component 202 includes an analysis engine 222. In some embodiments of the invention, the analysis engine 222 provides two interfaces: one for analyzing metric data and one for graphing metric data. In an exemplary embodiment of the invention, the interface for metric data analysis is named IDataAnalyzer. The public interface IDataAnalyzer includes a method Execute (DataSetList) that performs the actual analysis of the metric data in the DataSetList parameter and stores the analyzed data internally as another DataSetList. The public interface IDataAnalyzer also includes a property named CurrentDataSetList, which stores the analyzed data. Alternatively, a user may use the graph analyzer interface to provide some form of graphic analysis of metric data stored in the central metrics storage module 218. In an exemplary embodiment of the invention, the graph analyzer interface is named IGraphAnalyzer. The IGraphAnalyzer interface also includes an Execute (DataSetList) method. This method generates the graphs from the input metric data in the DataSetList, and stores the graphs internally as images. In addition, the IGraphAnalyzer interface includes a property named Graph, which allows a user to access the graphs generated from the Execute method. A DataSetList is an array of DataSet objects. Each DataSet object maps to a specific set of values for a given metric. That is, each DataSet object hosts a set of DataPoints and a set of attributes that go along with the DataPoints. Each DataPoint object maps to a specific value for a given metric. In other words, each DataPoint object holds attributes and data about a single data value of a metric. In some embodiments of the invention, an enumeration is provided to identify the type of the analyzer. For example, an analyzer can be a statistical data analyzer or a graphical analyzer. Another enumeration may be provided to identify the type of aggregation for a statistical analyzer. For example, the aggregation for a statistic analyzer can be mean, mode, range, or variance. Further, an enumeration may be provided to identify the type of graph for a graphical analyzer. For example, a graph can be a bar graph or a line graph. In some embodiments of the invention, the analysis component 220 may further comprise an analysis user interface 224 (hereinafter “analysis UI”). A user may analyze metric data in the central metrics storage module 218 through the analysis UI 224 instead of using the analysis engine 222. FIGS. 6A–6C illustrate an exemplary analysis UI 224. As shown in FIGS. 6A–6C, the analysis UI 224 provides three submenus: data set 602, analysis 618, and graph 624. The “data set” submenu 602 (FIG. 6A) provides access to various actions that can be performed on the data fetched from the central metrics storage module 218. For example, the data set menu 602 contains action items such as “new” 604, “load” 606, “save” 608, “copy rows” 610, and “append” 612. The “new” 604 action item allows a user to build a custom data set from existing data sets, for example, by either using the existing data sets as they are or merging various existing data sets. The “load” 606 action item allows a user to load a previously saved data set. The “save” 608 action item saves a currently selected data set containing the metric data from the central metrics storage module 218 to a permanent storage for offline analysis or reporting. The “copy rows” 610 action item allows a user to copy only a portion of data from an existing data set to a new data set. The “append” 612 action item allows a user to select the “columns” 614 option to merge one or more columns from one or more data sets into the currently selected data set. The “append” 612 action item also allows a user to select the “data sets” 616 option to merge one or more complete data sets into the currently selected data set. The analysis submenu 618 (FIG. 6B) enables a user to access analytical actions that can be performed on the data sets fetched from the central metrics storage module 218. For example, the analysis submenu 618 contains a link 620 to a statistical analyzer that allows a user to choose from all available statistical analyzers. A user may select one or more statistical analyzers to run on the current data set. The analysis submenu 618 may also provide a link 622 that allows a user to choose the kind of graphs the user wants to generate for the data set fetched from the central metrics storage module 218. The graph submenu 624 (FIG. 6C) enables a user to access functions that can be performed on the graphs that are generated from the data set fetched from the central metrics storage module 218. For example, a user can load 626 a graph of a data set. The user can also save 628 the current selected graph to permanent storage for offline analysis or reporting. FIG. 7 is a flow diagram illustrating an exemplary process 700 that collects and reports metric data. In essence, the process 700 reads information in a configuration file and/or one or more component specific plug-ins. The process 700 then collects metric data according to the settings specified in the configuration file and/or the component specific plug-ins. The process further verifies the collected metric data with one or more IPF client plug-ins. In the case that the metric data passed one or more criteria specified by one or more IPF plug-ins, the process 700 reports the metric data to a central metrics storage module. More specifically, upon receiving a request to collect data for one or more metrics, the process 700 reads a configuration file such as the configuration file 204 illustrated in FIG. 2. See block 702. If the component contains one or more data collection plug-ins, such as the data collection plug-in 504 illustrated in FIG. 5, the process 700 loads and initializes the one or more data collection plug-ins. See block 706. The process 700 then proceeds to collect data for one or more metrics. See block 708. If a metric is associated with one or more IPF client plug-ins, such as the IPF client plug-ins 506 illustrated in FIG. 5, the process 700 passes the collected metric data to check the results returned by the IPF client plug-ins. See block 710. The process 700 determines if the metric data passes the verification by one or more IPF client plug-ins. See decision block 712. If the metric data fails at least one IPF client plug-in verification, the process 700 proceeds to terminate the corresponding metric task and clean up the system, for example, by removing the metric data already collected. See block 714. The process 700 then terminates. If the answer to decision block 712 is YES, then the metric data successfully passes the verification by all involved IPF client plug-ins. The process 700 the proceeds to report the collected metric data to a central metrics storage module, such as the central metrics storage module 218 illustrated in FIG. 2. See block 716. The process 700 then waits for the next data collection interval. See block 718. As noted above in the description of the configuration UI 206, a data collection interval, i.e., report interval, is specified by a user through the configuration UI 206 or in the configuration file 204. While the preferred embodiment of the invention has been illustrated and described, it will be appreciated that various changes can be made therein without departing from the spirit and scope of the invention.
	summary
	summary
	claims	1. A composition to immobilize nuclear containing waste containing at least one radioactive element or alloy of uranium, graphite, magnesium, and aluminum, said composition comprising:at least one mineral phase forming element or compound that reacts with said at least one radioactive element or alloy; andat least one glass-forming element or compound to form a glass phase that incorporates waste radioisotopes and impurities that do not react with said least one mineral phase forming element or compound,wherein the composition immobilizes said nuclear containing waste in a solid wasteform,wherein the nuclear containing waste comprises a magnesium metal, carbonate or hydroxide, and the mineral phase is sufficient to convert the magnesium metal, carbonate or hydroxide to MgO, and/or the nuclear containing waste comprises a uranium metal and/or hydrated and carbonated uranium, and the mineral phase is sufficient to convert the uranium oxide. 2. The composition of claim 1, wherein nuclear containing waste comprises magnox sludges, clinoptilolite wastes, and combinations thereof. 3. The composition of claim 1, wherein the final wasteform has a density of greater than 90% theoretical density. 4. The composition of claim 1, wherein the magnesium has a magnesium-derived component with a particle size <250 μm. 5. The composition of claim 1, further comprising at least one additive to assist in the thermal treatment and consolidation of the final wasteform. 6. The composition of claim 5, wherein the at least one additive comprises a source of Ti, Si, P or Al, that react with the magnesium component to form stable phases. 7. The composition of claim 6, wherein the sources of Ti, Si, P or Al include titania, titanate minerals, alumina, phosphate silica, silicate minerals, silica sol and glass frit. 8. The composition of claim 6, wherein the stable phases comprise MgTiO3, Mg2TiO4, MgTi2O5, MgSiO3, Mg2SiO4, MgAl2O4, and combinations thereof. 9. The composition of claim 6, further comprising ternary multicomponent phases from other elements in the waste. 10. The composition of claim 9, wherein the ternary multicomponent phases comprise diopside (CaMgSi2O6) and perovskite (CaTiO3). 11. The composition of claim 1, wherein the wasteform comprises a mixture of MgO and other ceramic phases encapsulated in a protective glass plus ceramic phase matrix. 12. The composition of claim 1, wherein the wasteform comprises a uranium oxide or a compound selected from a titanate mineral: brannerite, pyrochlore, zirconolite encapsulated in a matrix. 13. The composition of claim 1, wherein the nuclear containing waste contains both coarse fraction and fine fractions of particulates in a ratio ranging from 10:90 to 90:10. 14. The composition of claim 13, wherein a majority of the coarse fractions have particulate sizes ranging from 200 μm to 6 mm and a majority of the fine fractions have particulate sizes ranging from 0.1 to less than 200 μm. 15. A method of treating nuclear containing waste, said method comprising:forming a slurry comprising nuclear containing waste containing a magnesium metal, carbonate or hydroxide and a composition comprising at least one mineral phase forming element or compound and at least one glass-forming element or compound, wherein the mineral phase is sufficient to convert the magnesium metal, carbonate or hydroxide to MgO, and/or the nuclear containing waste comprises a uranium metal and/or hydrated and carbonated uranium, and the mineral phase is sufficient to convert the uranium oxide;drying said slurry to form a dried product;calcining the dried product to form a calcined product;loading the calcined product into a canister;evacuating and sealing the metal canister; andthermally treating the calcined product in the canister to produce a dense wasteform product, which comprises a glass-ceramic. 16. The method of claim 15, further comprising mixing said slurry using at least one paddle type mixing, recirculating mixing, in-line mixing, turbulent slurry mixing or a combination thereof. 17. The method of claim 15, wherein drying the slurry is performed concurrently with a step that granulates the resulting product, said drying comprises at least one process selected from wipe or thin-film evaporation, rotary drying or conical mixer drying, spray drying, fluidized bed drying or flash drying. 18. The method of claim 15, wherein hot-isostatic pressing occurs at temperatures ranging from 900° C. and 1300° C. and pressures ranging from 5 MPa to 150 MPa. 19. The method of claim 15, wherein calcining occurs using a rotary calcination, vibratory calcination, fluidized bed calcination or a batch calcination method. 20. The method of claim 15, wherein calcining removes hydrogen and/or reactive water from the wasteform. 21. The method of claim 15, further comprising thermal processing to densify the product to greater than 90% theoretical density. 22. The method of claim 15, further comprising reacting a magnesium-derived component having a particle size less than 250 μm with at least one additive to assist in the thermal treatment and consolidation of the dense waste form product. 23. The method of claim 21, wherein said reacting comprises the intermixing of fine components of the waste with fine additives that provide sources of Ti, Si, P or Al, to form at least one stable phase. 24. The method of claim 23, wherein the sources of Ti, Si, P or Al include titania, alumina, phosphate silica, and glass frit. 25. The method of claim 23, wherein the at least one stable phase comprises MgTiO3, Mg2TiO4, MgTi2O5, MgSiO3, Mg2SiO4, MgAl2O4, and combinations thereof. 26. The method of claim 23, further comprising ternary multicomponent phases from other elements in the waste. 27. The method of claim 26, wherein the ternary multicomponent phases comprise diopside (CaMgSi2O6), perovskite (CaTiO3) and mixtures thereof. 28. The method of claim 15, wherein the dense wasteform product comprises a mixture of MgO and at least one other ceramic phase encapsulated in a protective glass plus ceramic phase matrix. 29. The method of claim 15, wherein the dense wasteform product comprises a uranium oxide and at least compound selected from a titanate mineral brannerite, pyrochlore, zirconolite encapsulated in a matrix. 30. The method of claim 15, wherein the at least one glass-forming element or compound comprises a zeolite. 31. The method of claim 30, wherein the zeolite comprises clinoptilolite, which is found in an amount to adsorb the volatile elements and isotopes thereof, to form a glass during processing that contains radioactive ions present in the waste, or both, such that their loss rate during thermal processing is reduced. 32. The method of claim 30, wherein the volatile elements and isotopes thereof comprise Cs. 33. The method of claim 30, wherein the zeolite reacts to form a glass during processing that contains radioactive ions present in the waste. 34. The method of claim 15, further comprising vibratory packing the calcined material prior to thermally treating. 35. The method of claim 34, wherein the canister is made of a metal, and thermally treating the metal material canister comprises hot isostatic pressing. 36. The method of claim 15, further comprising granulating the dried product.
049869574	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to zirconium based alloys suitable for use in nuclear reactor service, and in particular for use in claddings of fuel elements. 2. Description of Related Art Nuclear fuel element cladding serves several purposes and two primary purposes are: first, to prevent contact and chemical reactions between the nuclear fuel and the coolant or the moderator if a moderator is present; and second, to prevent the radioactive fission products, some of which are gases, from being released from the fuel into the coolant or the moderator. The failure of the cladding, i.e., a loss of the leak-proof seal, can contaminate the coolant or moderator and the associated systems with radioactive long-lived products to a degree which interferes with plant operation. Zirconium-based alloys have long been used in the cladding of fuel elements in nuclear reactors. A desirable combination is found in zirconium by virtue of its low thermal neutron cross-section and its generally acceptable level of resistance to corrosion in a boiling water reactor environment. Zircaloy 2, a zirconium alloy consisting of about 1.2 to 1.7 percent tin, 0.07 to 0.2 percent iron, 0.05 to 0.15 percent chromium, 0.03 to 0.08 percent nickel, up to 0.15 percent oxygen, and the balance zirconium, has enjoyed performance in reactor service, but also possesses some deficiencies that have prompted further research to find materials providing improved performance. For example, Zircaloy 2 cladding on fuel elements in nuclear reactors absorbs hydrogen while the reactor is operating. When the reactor is shut down and the cladding cools the Zircaloy 2 is embrittled by the absorbed hydrogen. Zircaloy 4 was one alloy developed as a result of further research to improve Zircaloy 2. Zircaloy 4 is similar to Zircaloy 2 but contains less nickel (0.007% max. wt. percent) and slightly more iron. Zircaloy 4 was developed as an improvement over Zircaloy 2 to reduce absorption of hydrogen in Zircaloy 2. Zircaloy 2 and Zircaloy 4 are herein referred to as the Zircaloy alloys or Zircaloy. The Zircaloy alloys are among the best corrosion resistant materials when tested in water at reactor operating temperatures, typically about 290.degree. C., but in the absence of radiation from the nuclear fission reaction. The corrosion rate in water at 290.degree. C. is very low and the corrosion product is a uniform, tightly adherent, black ZrO.sub.2 film. In actual service, however, the Zircaloy is irradiated and is also exposed to radiolysis products present in reactor water. The corrosion resistance properties of Zircaloy deteriorate under these conditions and the corrosion rate thereof is accelerated. Research efforts directed at improving the corrosion properties of the zirconium-based alloys have yielded some advances. Corrosion resistance has been enhanced in some instances through carefully controlled heat treatments of the alloys either prior to or subsequent to material fabrication. Added heat treatment cycles, however, generally increase the expense of making finished products, and in those instances where an installation requires welding to be performed, the area affected by the heat of the welding operation may not possess the same corrosion resistance characteristics as the remainder of the article. Variations in the alloying elements employed and the percentages of the alloying elements have also been propounded in an effort to address the deterioration in the corrosion-resistance of these alloys when they are irradiated. The deterioration under actual reactor conditions of the corrosion resistance properties of Zircaloy is not manifested in merely an increased uniform rate of corrosion. Rather, in addition to the black ZrO.sub.2 layer formed, a localized, or nodular corrosion phenomenon has been observed in some instances on Zircaloy tubing in boiling water reactors. In addition to producing an accelerated rate of corrosion, the corrosion product of the nodular corrosion reaction is a highly undesirable white ZrO.sub.2 bloom which is less adherent and lower in density than the black ZrO.sub.2 layer. The increased rate of corrosion caused by the nodular corrosion reaction will be likely to shorten the service life of the tube cladding, and also this nodular corrosion will have a detrimental effect on the efficient operation of the reactor. The white ZrO.sub.2, being less adherent, may be prone to spalling or flaking away from the tube into the reactor water. On the other hand, if the nodular corrosion product does not spall away, a decrease in heat transfer efficiency through the tube into the water is created when the nodular corrosion proliferates and the less dense white ZrO.sub.2 covers all or a large portion of a tube. Actual reactor conditions cannot be readily duplicated for normal laboratory research due to the impracticality of employing a radiation source to simulate the irradiation experienced in a reactor. Additionally, gaining data from actual use in reactor service is an extremely time consuming process. For this reason, there is no conclusory evidence in the prior art which explains the exact corrosion mechanism which produces the nodular corrosion. This limits, to some degree, the capability to ascertain whether other alloys will be susceptible to nodular corrosion before actually placing samples made from these alloys into reactors. Laboratory tests conducted under the conditions normally experienced in a reactor at approximately 300.degree. C. and 1000 psig in water, but absent radiation, will not produce a nodular corrosion product on Zircaloy alloys like that found in some instances on Zircaloy alloys which have been used in reactor service. However, if steam is used, with the temperature increased to over 500.degree. C. and the pressure raised to 1500 psig, a nodular corrosion product like that occasionally found on Zircaloy in reactor service can be produced on Zircaloy alloys in laboratory tests. Specimens of Zircaloy alloys which are annealed at 750.degree. C. for 48 hours are particularly susceptible to nodular corrosion under these test conditions. These annealed Zircaloy specimens will produce, in tests run for relatively short times, i.e. 24 hours, a degree of nodular corrosion comparable to that of Zircaloy tube cladding in actual reactor service that has been found to have nodular corrosion. At this higher temperature and pressure, a simulated nuclear reactor environment is provided which will allow researchers to determine the susceptibility of new alloys to nodular corrosion. With this test, a comparison between samples from new alloys and Zircaloy specimens tested under the same conditions can be made. To be considered as a suitable alternate or replacement for the Zircaloy alloys, any new alloy must not only be less susceptible than the Zircaloy alloys to nodular corrosion, but must maintain acceptable uniform corrosion rates, comparable to those of the Zircaloy alloys, to ensure sufficient service life. Zircaloy alloys have been used extensively as fuel rod cladding and are known to contain many desirable properties that alternate or replacement alloys must also contain. Zircaloy alloys have the desirable properties of a low neutron absorption cross section and at temperatures below 750.degree. F. are strong, ductile, extremely stable and as mentioned previously have excellent uniform corrosion resistance in water at reactor operating temperatures. Fuel element performance has revealed another problem with brittle splitting of nuclear fuel element cladding due to the combined interactions between the nuclear fuel, the cladding and the fission products produced during nuclear fission reactions. It has been discovered that this undesirable performance is due to localized mechanical stresses on the fuel cladding resulting from differential expansion and friction between the fuel and the cladding. Fission products are created in the nuclear fuel by the fission chain reaction during operation of a nuclear reactor, and these fission products are released from the nuclear fuel and are present at the cladding surface. These localized stresses and strains in the presence of specific fission products, such as iodine and cadmium, are capable of producing cladding failures by phenomena known as stress corrosion cracking or liquid metal embrittlement. SUMMARY OF THE INVENTION The present invention relates to corrosion resistant zirconium-based alloys and corrosion resistant nuclear fuel elements encased with cladding container tubing made from such corrosion resistant zirconium alloys. In one embodiment, a corrosion resistant first alloy consists essentially of by weight percent about 0.5 to 2.0 percent tin, about 0.24 to 0.40 percent of a solute composed of copper, nickel and iron wherein the copper is at least 0.05 percent, and the balance zirconium. In another embodiment, a corrosion resistant second alloy consists essentially of in weight percent about 0.5 to 2.0 percent tin, a solute composed of copper, iron and nickel so that each solute element is present in an amount from 0.05 to 0.20 percent, and the balance zirconium. In another embodiment, a corrosion resistant third alloy consists essentially of in weight percent about 0.5 to 2.0 percent tin, about 0.25 to 0.35 percent of a solute composed of copper and nickel wherein the copper is at least 0.05 percent, and the balance zirconium. These alloys provide increased resistance to nodular corrosion in high pressure and temperature steam testing, and will maintain acceptable uniform corrosion rates in water and steam tests. Corrosion-resistant nuclear fuel elements are provided by making elongated cladding containers from the first, second or third zirconium alloys described above. Improved corrosion resistant nuclear fuel elements are also made from composite cladding container tubing having a Zircaloy alloy tube with a surface layer metallurgically bonded on the outside of the Zircaloy tube. The surface layer being about 5 to 20 percent of the thickness of the Zircaloy tube and consisting essentially of the first, second, or third zirconium alloy described above. The surface layer is a protective shield thick enough to prevent nodular corrosive attack on the Zircaloy tube. Another nuclear fuel element is made from a composite cladding container that is resistant to nodular corrosion, stress corrosion cracking and liquid metal embrittlement. An elongated composite cladding container is made from a Zircaloy alloy tube having a corrosion resistant surface layer metallurgically bonded to the outside surface and an inner barrier layer of zirconium metallurgically bonded on the inside of the alloy tube. The inner barrier being about 1 to 30 percent of the thickness of the Zircaloy tube and comprised of moderate purity zirconium such as sponge zirconium. The outer surface layer being about 5 to 20 percent of the thickness of the Zircaloy tube, and consisting essentially of the first, second or third zirconium alloy described above. Cladding container tubing is manufactured by heating an extrusion billet of the first, second or third zirconium alloy described above to about 590.degree. to 650.degree. C., extruding the billet into tube shell followed by standard tube reduction and subsequent heat treatments at about 570.degree. to 590.degree. C. to achieve desired tube dimensions and mechanical properties. The standard tube reduction process of zirconium alloy tubing used in nuclear fuel elements is pilger-rolling. Pilger-rolling is a tube reduction process using traveling, rotating dies on the outer tube surface to forge the tube over a stationary mandrel die inside the tube. Composite cladding containers are manufactured by starting with a tube blank made from a Zircaloy alloy, and an outer tube placed on this tube blank. The outer tube is composed of the first, second or third zirconium alloys described above. This composite tube is then heated to a temperature in the range of 590.degree. to 650.degree. C. and is extruded. In the process, a metallurgical bond between the two zirconium alloys results. Subsequent tube reduction and heat treatments between 570.degree. to 590.degree. C. are performed to achieve the desired tube dimensions and mechanical properties. The outer tube is of at least a thickness so that after tube reduction it is about 5 to 20 percent of the thickness of the Zircaloy tube. Another composite cladding container is manufactured by starting with a tube blank of a Zircaloy alloy and an outer tube is placed on the tube blank. The outer tube is made from the first, second or third zirconium alloys described above. A hollow collar of a metal barrier is placed inside the tube blank. The metal barrier is comprised of moderate purity zirconium such as sponge zirconium. The composite tubing is heated to 590.degree. to 650.degree. C. and extruded to form a metallurgical bond between the outside surface layer and the tube blank, and between the inner metal barrier and the tube blank. Extrusion is followed by tube reduction and subsequent heat treatments between 570.degree. C. and 590.degree. C. to develop the desired tube dimensions and mechanical properties. The outer tube and hollow collar are of at least a thickness so that after tube reduction the outer layer is about 5 to 20% of the thickness of the Zircaloy tube and the inner barrier is about 1 to 30% of the Zircaloy tube thickness. The cladding containers and the composite cladding containers enclose a nuclear fuel material, leaving a gap between the fuel and the cladding. In the composite cladding containers having a barrier layer, the barrier layer shields the alloy tube from the nuclear fuel material held in the cladding as well as shielding the alloy tube from fission products and gases. Because of its purity the liner remains soft during irradiation and minimizes localized strain inside the nuclear fuel element, thus serving to protect the alloy tube from stress corrosion cracking or liquid metal embrittlement. The first, second or third zirconium alloys and the barrier layer do not introduce any significant neutron capture penalites, heat transfer penalites, or material incompatibility problems for the nuclear fuel elements of this invention. DETAILED DESCRIPTION OF THE INVENTION The alloys of the present invention have demonstrated adequate resistance to uniform corrosion to be considered for nuclear reactor service, and the corrosion resistance is generally comparable to the excellent uniform corrosion resistance possessed by the Zircaloy alloys. The alloys of the present invention also demonstrate improved resistance to nodular corrosion. The addition of tin to zirconium has been practiced in the art prior to this invention, as evidenced by Zircaloy and other known zirconium-based alloys. The presence of tin, which stabilizes the alpha-form of zirconium, primarily contributes to the strength of the alloy, although there is some improvement in uniform corrosion resistance attributable to tin. It has been determined that concentration levels below about 0.5% by weight of tin will yield an alloy in which the uniform corrosion rate in water is unacceptably high. Additionally, concentration levels in excess of about 2.0% by weight of tin will yield an alloy which demonstrates an unacceptable level of accelerated corrosion in laboratory steam tests. The alloys of the present invention have a tin content within the range of from about 0.5 to about 2.0 percent by weight, preferably from 1.0 to about 1.5 percent by weight, and most preferably contain about 1.5% by weight of tin. The alloys of the present invention further contain certain additional alloying elements, termed collectively herein as a solute portion of the alloy. The solute portion of the alloys of the present invention differs from the additional alloying elements found in the Zircaloy alloys, and is primarily responsible for the comparative increase in resistance to nodular corrosion. Conventional impurities will also be present in these alloys. It should be noted that the alloys of the present invention will also optionally contain from about 0.09 to 0.16 weight percent of oxygen. Most commercial grade sponge zirconium which would be used in making alloys such as the ones in the present invention will contain small amounts of oxygen, roughly on the order of about 800-1300 parts per million. In some instances, it will be desirable to increase the concentration of oxygen in the alloy. Adding oxygen is one way to increase room temperature yield strength. Thus, the alloys of the present invention may be produced with or without the additional oxygen, as this will have little or no effect on the corrosion resistance of the alloys. There are several parameters which should be considered in choosing alloying elements for the candidate zirconium based alloys to be used for fuel cladding in boiling water reactor service. The thermal neutron cross-section of the element should be relatively low to permit products of the fission reaction to easily pass through the fuel cladding, thereby allowing the boiling water reactor to operate as efficiently as possible. The cost of the material should be taken into account, and must not be prohibitively high. The ease or difficulty with which an alloy containing the element or elements and zirconium can be produced must also be considered. It is further desired that the element or elements will enhance the corrosion resistance properties of the zirconium under actual or simulated boiling water reactor conditions. The thermal neutron cross-section of an element is generally a known property of the element if it has ever come under consideration for use in a nuclear reactor. The costs of the materials can be ascertained from historic price data, with extrapolation if required. The alloying process of the alloys of the present invention is similar to conventional methods for alloying zirconium and thus ease of making alloy additions is fairly predictable. The alloying is accomplished preferably by arc melting a zirconium billet having a suitable amount of the alloying metals encased in a hollow portion of the billet. This molten metal is then cast as an alloy billet, which will then be subjected to finishing processes to produce final shapes. Generally, of the parameters discussed above, the most difficult to predict is whether the alloying element will contribute to the enhancement of corrosion resistance. In the present invention, zirconium.-based alloys have been discovered that perform substantially better than Zircaloy 2 in tests conducted to determine resistance to nodular corrosion. These alloys also perform well in tests for determining resistance to uniform corrosion. A first alloy consists essentially of by weight percent 0.5 to 2.0 percent tin, about 0.24 to 0.40 percent of a solute composed of copper, nickel and iron, wherein the copper is at least 0.05 percent, and the balance zirconium. A second alloy consists essentially of by weight percent about 0.5 to 2.0 percent tin, a solute composed of copper, iron and nickel so that each solute element is present in an amount from 0.05 to 0.20 percent, and the balance zirconium. A third alloy consists essentially of by weight percent about 0.5 to 2.0 percent tin, about 0.25 to 0.35 weight percent of a solute composed of copper and nickel wherein the copper is at least 0.05 percent, and the balance zirconium. Solute elements copper, nickel and iron possess the low thermal neutron cross-section, low cost, ease of alloying, and corrosion resistance properties that are desirable in zirconium based alloys. Tests for both uniform corrosion resistance and nodular corrosion have been conducted on alloys of the present invention. These tests have shown that a dramatic decrease in susceptibility to nodular corrosion can be attained in an alloy which is relatively insensitive to heat treatment while retaining essentially the same uniform corrosion resistance of a Zircaloy 2 alloy. Solute concentrations ranging from as low as 0.24 weight percent to as high as 0.40 weight percent have been tested and have been shown to exhibit superior resistance to nodular corrosion, compared to the performance of Zircaloy 2. Alloys having copper and nickel as solutes had greatly improved resistance to nodular corrosion when they were given the 750.degree. C./48 hour anneal that sensitizes Zircaloy 2 to nodular corrosion. Zirconium alloy tubing is heat treated several times during tube production; therefore, the zirconium alloys containing solutes copper and nickel will provide improved nodular corrosion resistance when properly heat treated during tube production. The corrosion-resistant fuel elements of this invention are shown by referring now more particularly to FIG. 1, where there is shown a partially cutaway sectional view of a nuclear fuel assembly 10. This fuel assembly consists of a tubular flow channel 11 of generally square cross section provided at its upper end with lifting bale 12 and at its lower end with a nose piece (not shown due to the lower portion of assembly 10 being omitted). The upper end of channel 11 is open at 13 and the lower end of the nose piece is provided with coolant flow openings. An array of fuel elements or rods 14 is enclosed in a channel 11 and supported therein by means of upper end plate 15 and a lower end plate (not shown due to the lower portion being omitted). The liquid coolant ordinarily enters through the openings in the lower end of the nose piece, passes upwardly around fuel elements 14, and discharges at upper outlet 13 in a partially vaporized condition for boiling reactors or in an unvaporized condition for pressurized reactors at an elevated temperature. The nuclear fuel elements or rods 14 are sealed at their ends by means of end plugs 18 welded to the cladding 17, which may include studs 19 to facilitate the mounting of the fuel rod in the assembly. A void space or plenum 20 is provided at one end of the element to permit longitudinal expansion of the fuel material and accumulation of gases released from the fuel material. A nuclear fuel material retainer means 24 in the form of a helical member is positioned within space 20 to provide restraint against the axial movement of the pellet column, especially during handling and transportation of the fuel element. The fuel element is designed to provide an excellent thermal contact between the cladding and the fuel material, a minimum of parasitic neutron absorption, and resistance to bowing and vibration which is occasionally caused by flow of the coolant at high velocity. A nuclear fuel element or rod 14 is shown in a partial section in FIG. 1 constructed according to the teachings of this invention. The fuel element includes a core or central cylindrical portion of nuclear fuel material 16, here shown as a plurality of fuel pellets of fissionable or fertile material positioned within a structural cladding or container 17. In some cases, the fuel pellets may be of various shapes, such as cylindrical pellets or spheres, and in other cases, different fuel forms such as particulate fuel may be used. The physical form of the fuel is immaterial to this invention. Various nuclear fuel materials may be used, including uranium compounds, plutonium compounds, thorium compounds and mixtures thereof. A preferred fuel is uranium dioxide or a mixture comprising uranium dioxide and plutonium dioxide. Referring now to FIG. 2, the nuclear fuel material 16 forming the central core of the fuel element 14 is surrounded by a cladding 17. The cladding container encloses the core so as to leave a gap between the core and the cladding container during use in a nuclear reactor. The cladding is comprised of a corrosion-resistant zirconium alloy tube 21. The alloy tube 21 is made from either the first, second or third zirconium alloys described above. It should be noted that the first, second or third zirconium alloys described herein will also optionally contain from about 0.09 to 0.16 weight percent of oxygen. Most commercial grade sponge zirconium which would be used in making alloys such as the ones in the present invention will contain small amounts of oxygen, roughly on the order of about 800-1300 parts per million. In some instances, it will be desirable to increase the concentration of oxygen in the alloy. Adding oxygen is one way to increase room temperature yield strength. Thus, the alloys of the present invention may be produced with or without the additional oxygen, as this will have little or no effect on the corrosion resistance of the alloys. Another embodiment of this invention is shown by referring to FIG. 3. The nuclear fuel material 16 forming the central core of the fuel element 14 is surrounded by a composite cladding 17. The composite cladding container encloses the core so as to leave a gap 23 between the core and the cladding container during use in a nuclear reactor. The composite cladding is comprised of a zirconium alloy tube 21 made from either the first, second or third zirconium alloy described above. The alloy tube has bonded on the inside surface thereof a metal barrier 22 so that the metal barrier forms a shield between the alloy tube 21 and the nuclear fuel material held in the cladding The metal barrier forms about 1 to about 30 percent of the thickness of the cladding, and is comprised of a low neutron absorption material, namely, moderate purity zirconium. One moderate purity zirconium is sponge zirconium. The metal barrier 22 protects the alloy tube portion of the cladding from contact and reaction with gasses and fission products from the nuclear fuel, and prevents the occurrence of localized stress and strain. The content of the metal barrier of moderate purity zirconium is important and serves to impart special properties to the metal barrier Generally, there is at least about 1,000 parts per million (ppm) by weight and less than about 5,000 ppm impurities in the material of the metal barrier and preferably less than about 4,200 ppm. Of these, oxygen is kept within the range of about 200 to about 1,200 ppm. All other impurities are within the normal range for commercial, reactor-grade sponge zirconium. In another embodiment of this invention, a corrosion-resistant nuclear fuel element is shown by referring to FIG. 4. The nuclear fuel material 16 forming the central core of fuel element 14 is surrounded by a composite cladding 17. The composite cladding container encloses the core so as to leave a gap 23 between the core and the cladding container during use in a nuclear reactor. The composite cladding is comprised of a zirconium alloy tube 30 made from a Zircaloy alloy. The alloy tube has bonded on the outside surface thereof a metal layer 32 so that the metal layer forms a corrosion protective shield over the alloy tube. The outer metal layer is about 5 to 20 percent of the thickness of the alloy tube and is comprised of either the first, second or third zirconium alloy described above. The outer metal layer protects the Zircaloy alloy tube portion of the cladding from nodular corrosion. Another improved nuclear fuel element is shown by referring to FIG. 5. The nuclear fuel material 16 forming the central core of fuel element 14 is surrounded by a composite cladding 17. The composite cladding container encloses the core so as to leave a gap 23 between the core and the cladding container during use in a nuclear reactor. The composite cladding is comprised of a Zircaloy alloy tube 30. The alloy tube has bonded on the inside surface thereof a metal barrier 22 so that the metal barrier forms a shield between the alloy tube 30 and the nuclear fuel material held in the cladding. The metal barrier is about 1 to about 30 percent of the thickness of the alloy tube and is comprised of a low neutron absorption material, namely, moderate purity zirconium as described above. The metal barrier 22 protects the alloy tube portion of the cladding from contact and reaction with gases and fission products from the nuclear fuel, and prevents the occurrence of localized stress and strain. An outer surface layer is bonded on the outside surface of the alloy tube 30. The outer metal layer is about 5 to 20 percent of the thickness of the alloy tube and is comprised of either the first, second or third zirconium alloy described above. The outer metal layer protects the Zircaloy alloy tube portion of the cladding from nodular corrosion. Sponge zirconium metal forming the metal barrier in the composite cladding is highly resistant to radiation hardening, and this enables the metal barrier after prolonged irradiation to maintain desirable structural properties such as yield strength and hardness at levels considerably lower than those of conventional zirconium alloys. In effect, the metal barrier does not harden as much as conventional zirconium alloys when subjected to irradiation, and this together with its initially low yield strength enables the metal barrier to deform plastically and relieve pellet-induced stresses in the fuel element during power transients. Pellet induced stresses in the fuel element can be brought about, for example, by swelling of the pellets of nuclear fuel at reactor operating temperatures (300.degree. C. to 350.degree. C.) so that the pellet comes into contact with the cladding. It has further been discovered that a metal barrier of sponge zirconium of the order preferably about 5 to 15 percent of the thickness of the cladding and a particularly preferred thickness of 10 percent of the cladding bonded to the alloy tube of a zirconium alloy provides stress reduction and a barrier effect sufficient to prevent failures in the composite cladding. The corrosion resistant nuclear fuel rod cladding used in the nuclear fuel elements of this invention can be fabricated from a billet comprised of a zirconium alloy made from either the first, second or third zirconium alloy described above. The billet is heated to 590.degree. to 650.degree. C. and extruded. The extruded tubing is then subjected to a process involving conventional tube reduction until the desired size of tubing is achieved. In another method, a hollow collar of the sponge zirconium selected to be the metal barrier is inserted into a hollow billet of the first, second or third zirconium alloy described above. The assembly is heated to 590.degree. to 650.degree. C. and extruded. The extruded tubing is then subjected to a process involving conventional tube reduction until the desired size of cladding is achieved. In another method, a tube blank is made from a Zircaloy alloy and an outer tube is placed on this tube blank. The outer tube is composed of the first, second or third zirconium alloy described above. This assembly is then heated to a temperature in the range of 590.degree. to 650.degree. C. and is extruded. The extruded tubing is then subjected to a process involving conventional tube reduction until the desired size of tubing is achieved. In another method, a tube blank is made from a Zircaloy alloy and an outer tube is placed on this tube blank. The outer tube is composed of the first, second or third zirconium alloy described above. A hollow collar of the sponge zirconium selected to be the metal barrier is inserted into the tube blank. The assembly is heated to a temperature in the range of 590.degree. to 650.degree. C. and is extruded. The extruded tubing is then subjected to a process involving conventional tube reduction until the desired size of tubing is achieved. Intermediate and final anneals are used during the tube reduction processes described above. Anneals range between 570.degree. to 590.degree. C. The invention includes a method of producing a nuclear fuel element comprising making a cladding or a composite cladding container comprised of a zirconium alloy, or a zirconium alloy and a barrier layer, or a Zircaloy alloy and a surface layer, or a Zircaloy alloy and an outer surface layer and an inner barrier layer. The container is open at one end and filled with a core of nuclear fuel material leaving a gap between the core and the container and leaving a cavity at the open end. A nuclear fuel material retaining means is inserted into the cavity and an enclosure is applied to the open end of the container, leaving the cavity in communication with the nuclear fuel. The end of the clad container is then bonded to the enclosure to form a tight seal therebetween. The present invention offers several advantages promoting a long operating life for a nuclear fuel element. A greater resistance to nodular corrosion protects the strength and integrity of the cladding. On cladding having a barrier layer, the reduction of chemical interaction on the cladding, the minimization of localized stress on the zirconium alloy tube portion of the cladding, and the minimization of stress corrosion and strain corrosion on the zirconium alloy tube portion of the cladding, all reduce the probability of a splitting failure occurring in the zirconium alloy tube. The invention further reduces expansion or swelling of the nuclear fuel into direct contact with the zirconium alloy tube, and this reduces the occurrence of localized stress on the zirconium alloy tube, initiation or acceleration of stress corrosion of the alloy tube and bonding of the nuclear fuel to the alloy tube. An important property of the composite cladding of this invention is that the foregoing improvements are achieved with no substantial additional neutron absorption penalty. Further, the composite cladding has a very small heat transfer penalty in that there is no thermal barrier to transfer of heat, such as results in a situation where a separate foil or a liner is inserted in a fuel element. Also, the composite cladding of this invention is inspectable by conventional non-destructive testing methods during various stages of fabrication and operation. The following examples are offered to further illustrate the improved nodular corrosion resistance of the alloys used in this invention.
	description	This application claims priority to and the benefit of Korean Patent Application No. 10-2013-0100582, filed on Aug. 23, 2013, the disclosure of which is incorporated herein by reference in its entirety. 1. Field The present disclosure relates to an in-vessel control rod drive mechanism and a nuclear reactor with the same, and more particularly, to an in-vessel control rod drive mechanism and a nuclear reactor with the same, in which a control rod drive mechanism letting a control rod regulate the reactivity of a reactor core is installed in a nuclear reactor. 2. Discussion of Related Art According to installation positions of major components (e.g., a steam generator, a pressurizer, a reactor coolant pump, and so on), a nuclear reactor is classified into a loop type reactor (e.g., a conventional pressurized water reactor of Korea) where the major components are installed outside the reactor vessel, and an integral type reactor (e.g., a system-integrated modular advanced reactor (SMART)) where the major components are installed inside a reactor vessel. And the nuclear reactor may be also classified into a pressurized water reactor in which boiling of a coolant is prevented, and a boiling water reactor in which the boiling of the coolant is allowed. The nuclear reactor uses a control rod to control reactivity (a nuclear reaction, fission) of a reactor core. The control rod is operated by a control rod drive mechanism (CRDM). One example of a control rod drive mechanism is disclosed in “Control rod drive mechanism installed in reactor pressure vessel”, Technical Situation Analysis Report of Korea Atomic Energy Research Institute (KAERI) (KAERI/AR-809/2008) published in September, 2008. Meanwhile, if a CRDM is provided in a reactor vessel, the CRDM should be operated in a high temperature, high pressure, and highly-radioactive primary coolant. Therefore, it should be preceded by technical development of key components of the in-vessel control rod drive mechanism. Also in order to realize a boron-free reactor core, a miniaturization technology for the in-vessel control rod drive mechanism should be developed at the same time. One aspect of the present invention is directed to an in-vessel control rod drive mechanism and a nuclear reactor with the same, in which a control rod drive mechanism is alternately arranged at an upper or lower side of a reactor core and installed inside the nuclear reactor. Another aspect of the present invention is also directed to an in-vessel control rod drive mechanism and a nuclear reactor with the same, in which a difficulty in miniaturizing the in-vessel control rod drive mechanism may be relieved by the alternate arrangement, and a boron-free reactor core may be also easily realized. One aspect of the present invention provides an in-vessel control rod drive mechanism including a control rod drive mechanism for regulating and a control rod drive mechanism for shutdown provided at an upper or lower space of a reactor core to insert or withdraw a regulating rod and a shutdown rod into/from the reactor core based on an operation state of a nuclear reactor, wherein the control rod drive mechanism for regulating and the control rod drive mechanism for shutdown are alternately arranged. Another aspect of the present invention provides a nuclear reactor including an in-vessel control rod drive mechanism according to the present invention, which is provided inside a pressure vessel of the reactor, and inserts or withdraws a regulating rod and a shutdown rod into/from a reactor core based on an operation state of the nuclear reactor. According to the nuclear reactor with the in-vessel control rod drive mechanism as described above, the control rod drive mechanism for regulating and the control rod drive mechanism for shutdown can be divided and alternately arranged at the upper and lower sides of the reactor core, can be divided and alternately arranged at the upper side of the reactor core, or can be divided and alternately arranged at the lower side of the reactor core. Therefore, according to embodiments of the present invention, the control rod drive mechanism for regulating and the control rod drive mechanism for shutdown can be respectively installed to have a different structure from each other. Even when a width or a diameter of the control rod drive mechanism is greater than a width of the fuel assembly, a plurality of control rod drive mechanisms can be installed and it is possible to accommodate more CRDMs than internal CRDMs of convention arrange. The control rod drive mechanism for regulating and the control rod drive mechanism for shutdown can be easily arranged without the significant obstruction of flow through the reactor core. The upper structure of the reactor core and the penetrating structure of the pressure vessel can be simplified, and the fluid passage resistance of the nuclear reactor cooling system can be reduced. According to embodiments of the present invention, the plurality of control rod drive mechanisms can be installed, and thus a sufficient reactor core shutdown margin can be provided. According to embodiments of the present invention, when the control rod drive mechanism for shutdown is provided at the lower side of the reactor core, the lower portion of the shutdown rod can be used as the burnable poison rod, and thus the boron-free reactor core can be easily realizable. According to embodiments of the present invention, it is possible to accommodate more CRDMs than internal CRDMs of convention arrange, and the rod worth per each control rod drive mechanism can be reduced, and thus an influence of the malfunction of the control rod can be minimized. According to embodiments of the present invention, a space of containment can be minimized due to the installation of the in-vessel control rod drive mechanism, and thus the rod ejection accident can be prevented, and the loss-of-coolant accident can be reduced. Hereinafter, an in-vessel control rod drive mechanism and a nuclear reactor with the same according to the embodiment of the present invention will be described in detail with reference to accompanying drawings. In the present disclosure, although in the singular number, a noun is construed as in the plural number, except as distinctively expressed in context. Generally, the CRDM which is applied to the nuclear reactor is installed outside the reactor vessel. That is, a top-inserted CRDM which is installed at a top of the nuclear reactor is used in a pressurized water reactor, and a bottom-inserted CRDM which is installed at a bottom of the nuclear reactor is used in a boiling water reactor. An external CRDM installed outside the reactor vessel is installed at a nozzle passing through the reactor vessel, and thus the CRDM forms part of a reactor pressure boundary together with a nozzle. Therefore, there is a possibility that the external CRDM may be damaged, and thus there is a risk that a loss-of-coolant accident or a rod ejection accident may occur. For addressing the foregoing possibility, an in-vessel CRDM is provided. The in-vessel CRDM is capable of fundamentally excluding the accidents and improving safety of a nuclear power plant, installing a plurality of control rods to realize a boron-free reactor core and to minimize related equipment thereof, and installing the CRDM inside the reactor vessel in an effort to simplify a reactor vessel penetration design and to reduce an amount of radioactive waste, have been developed. An in-vessel control rod drive mechanism according to embodiments of the present invention is arranged inside a nuclear reactor to prevent a rod ejection accident, to easily realize a boron-free reactor core, to enhance use of a space at an upper or lower side of a reactor vessel, and to miniaturize a containment. In particular, the in-vessel control rod drive mechanism according to embodiments of the present invention may be divided into upper and lower portions and alternately arranged at upper and lower sides of the reactor core, may be divided into upper and lower portions and alternately arranged at the upper side of the reactor core, or may be divided into upper and lower portions and alternately arranged at the lower side of the reactor core. As described above, even when a width or a diameter of the in-vessel control rod drive mechanism according to embodiments of the present invention is greater than a width of a fuel assembly, the in-vessel control rod drive mechanism may be disposed at each fuel assembly to sufficiently secure a means for restricting reactivity of the reactor core, and thus stability of the nuclear reactor may be enhanced, and the boron-free reactor core may be easily realized. Embodiments of the present invention may allow a primary coolant to flow with small resistance at the upper or lower side of the reactor core, may relieve a difficulty of flow uniformity at the reactor core, and thus may solve a problem that the fluid flow at the upper or lower side of the reactor core is disturbed by the control rod drive mechanism and a fluid passage resistance is increased. Hereinafter, the nuclear reactor with the in-vessel control rod drive mechanism according to embodiments of the present invention will be described fully. [First Embodiment] FIG. 1 is a configuration view of a nuclear reactor with an in-vessel control rod drive mechanism according to a first embodiment of the present invention, and FIG. 2 is a partially enlarged view of the reactor illustrated in FIG. 1. For convenience of explanation in the embodiment, an integral type reactor (hereinafter, called as a “reactor”) will be used. However, the present invention is not always limited thereto. That is, embodiments of the present invention may be applied to a loop type reactor as well as the integral type reactor. As illustrated in FIG. 1, the nuclear reactor 10 may include a pressure vessel 11, a fuel assembly 13 installed inside the pressure vessel 11 and having therein a reactor core 12 generating heat energy using a nuclear reaction of a nuclear fuel, a control rod 20 regulating reactivity of the reactor core 12, a steam generators 14 configured to transfer the heat energy generated from the reactor core 12 to a secondary coolant and thus to generate steam, a pressurizer 15 configured to control a pressure of a primary cooling system of the nuclear reactor 10, and a reactor coolant pump 16 configured to circulate a primary coolant. Of course, in the case of the loop type reactor, the steam generator 14, the pressurizer 15, and the reactor coolant pump 16 may be disposed outside the pressure vessel 11. FIG. 3 is a partially enlarged cross-sectional view of the reactor core and the control rod illustrated in FIG. 2, FIG. 4 is a cross-sectional view taken along line A-A′ of FIG. 2, and FIG. 5 is a cross-sectional view taken along line B-B′ of FIG. 2. As illustrated in FIGS. 2 and 3, the control rod 20 is classified into a regulating rod 21 which regulates the degree of insertion when the nuclear reactor 10 is in startup or normal operation mode, and regulates the reactivity of the reactor core 12, and a shutdown rod 22 which inserts negative reactivity when the nuclear reactor 10 is in shutdown mode, and shutdowns the nuclear reactor 10. FIG. 3 shows that the control rod 20 is installed at all of the fuel assembly 13. However, the present invention is not limited thereto, and the control rod 20 may not be installed at part of the fuel assembly 13 according to characteristics required in the nuclear reactor. The regulating rod 21 may include a control rod for startup 23 and a control rod for normal operation 24. The control rod for startup 23 may perform a function of compensating a change in the reactivity between a zero power mode with high temperature and a full power mode with high temperature, and the control rod for normal operation 24 may perform a function of compensating a change in a degree of fuel burn-up according to the elapse of operation time. The shutdown rod 22 may compensate the change in the reactivity between a zero power mode with low temperature and the zero power mode with high temperature. Meanwhile, when it is necessary to urgently shutdown the nuclear reactor 10, power is shut off, and the regulating rod 21 and the shutdown rod 22 are urgently inserted into the reactor core 12 by force of gravity, force of pressure, force of a spring, or the like, and perform a function of shutting down the nuclear reactor 10. Generally, in the integral type reactor, it is easy to provide an available space for installing a control rod drive mechanism 30 at upper and lower sides of the reactor core 12. Therefore, in the embodiment, as illustrated in FIG. 2, the in-vessel control rod drive mechanism 30 is alternately arranged at the upper and lower sides of the reactor core 12. More specifically, as illustrated in FIGS. 4 and 5, the in-vessel control rod drive mechanism 30 according to the first embodiment of the present invention may include a control rod drive mechanism for regulating 31 provided at the upper side of the reactor core 12 and configured to insert or withdraw the regulating rod 21 into/from the reactor core 12 according to an operation state of the nuclear reactor 10, and a control rod drive mechanism for shutdown 32 provided at the lower side of the reactor core 12 and configured to insert or withdraw the shutdown rod 22 into/from the reactor core 12 according to the operation state of the nuclear reactor 10. As described above, embodiments of the present invention may install the control rod drive mechanism for regulating 31 and the control rod drive mechanism for shutdown 32 at the upper and lower sides of the reactor core 12 without significant obstruction of flow through the reactor core 12. The control rod drive mechanism for shutdown 32 may be fabricated to have a relatively simple structure, compared with the control rod drive mechanism for regulating 31 in which precise control is required. Therefore, in the embodiment, the control rod drive mechanism for shutdown 32 is installed at the lower side of the reactor core 12. As the control rod drive mechanism for shutdown 32 is installed at the lower side of the reactor core 12, a lower portion of the shutdown rod 22 may be used as a burnable poison rod, and thus a boron-free reactor core may be easily realizable. Meanwhile, FIGS. 6 to 8 are views illustrating installation states of the nuclear fuel assembly and the control rod. As illustrated in FIG. 6, the control rod 20 may be fabricated to have a larger diameter than that of a fuel rod 17 arranged in the fuel assembly 13, and then installed to be spaced a predetermined distance. Alternatively, as illustrated in FIG. 7, the control rod 20 may be fabricated to have a similar diameter to that of the fuel rod 17 arranged in the fuel assembly 13, and then installed so that all of the control rods 20 form a certain shape. Or, as illustrated in FIG. 8, the control rod 20 may have an approximately ‘+’-shaped cross section, and may be installed among the fuel assembly 13 having an approximately rectangular plate-shaped cross section. As described above, embodiments of the present invention may change a position of the control rod according to standards and shapes of the fuel assembly, the fuel rod, and the control rod, and may effectively install the control rod. Next, an operation method of the nuclear reactor with the in-vessel control rod drive mechanism according to the embodiment of the present invention will be described in detail with reference to FIGS. 9 to 14. FIG. 9 is a flowchart illustrating each process in the operation method of the nuclear reactor with the in-vessel control rod drive mechanism according to the embodiment of the present invention, and FIGS. 10 to 13 are views illustrating operation states according to the operation method of the nuclear reactor with the in-vessel control rod drive mechanism illustrated in FIG. 9. In FIG. 9, when the nuclear reactor 10 is in startup mode, the nuclear reactor 10 and a related system are preheated to be operated in normal operation mode, and the in-vessel control rod drive mechanism 30 withdraws the control rods for shutdown, startup, and normal operation 22, 23, and 24, in turn, from the reactor core 12, and startups the reactor core 12. As illustrated in FIG. 10, all of the control rods for shutdown, startup, and normal operation 22, 23, and 24 are inserted into the reactor core 12 before starting up the nuclear reactor 10. In a process S10 of startup processes of the nuclear reactor 10, as illustrated in FIG. 11, the control rod drive mechanism for shutdown 32 connected to the shutdown rod 22 withdraws the shutdown rod 22 from the reactor core 12. At this time, the control rod drive mechanism for shutdown 32 may be operated to withdraw the shutdown rod 22 to the upper side of the reactor core 12 using driving force of a motor, force of pressure, force of gravity, or restored force of a spring. Of course, the control rod drive mechanism for shutdown 32 may be operated to withdraw the shutdown rod 22 to the lower side of the reactor core 12. In a process S12, as illustrated in FIG. 12, the control rod drive mechanism for regulating 31 connected to the control rod for startup 23 withdraws the control rod for startup 23 from the reactor core 12. At this time, the control rod drive mechanism for regulating 31 may be operated to withdraw the control rod for startup 23 to the upper side of the reactor core 12 using the driving force of the motor, force of pressure, the force of gravity, or the restored force of the spring. In a process S14, as illustrated in FIG. 13, the control rod drive mechanism for regulating 31 connected to the control rod for normal operation 24 withdraws the control rod for normal operation 24 from the reactor core 12. At this time, the control rod drive mechanism for regulating 31 may be operated to withdraw the control rod for normal operation 24 to the upper side of the reactor core 12 using the driving force of the motor, the force of pressure, the force of gravity, or the restored force of the spring. Meanwhile, in the process S12 of withdrawing the control rod for startup 23 and the process S14 of withdrawing the control rod for normal operation 24, the control rod for normal operation 24 is not always withdrawn after the control rod for startup 23 is withdrawn. Embodiments of the present invention may be modified such that the control rod for startup 23 and the control rod for normal operation 24 are alternately withdrawn according to the characteristics required in the nuclear reactor. If the control rods for shutdown, startup, and normal operation 22, 23, and 24 are withdrawn in turn through the processes S10 to S14, the reactor core 12 startups, and the nuclear reactor 10 moves to normal operation mode (S16). At this time, the control rod drive mechanism for regulating 31 may control the degree of insertion of the control rods for startup and normal operation 23 and 24, and thus may regulate the reactivity of the reactor core 12. Therefore, the heat energy is generated in the reactor core 12 due to the nuclear reaction of the nuclear fuel, and it is transferred to the steam generator 14. The steam generators 14 convert the secondary coolant to steam with the transferred energy, and then supply the steam to a turbine (not shown). Meanwhile, when the operation of the nuclear reactor 10 is ended, the control rods for shutdown, startup, and normal operation 22, 23, and 24 may be controlled to be inserted into the reactor core 12, such that the nuclear reactor 10 is shutdown. Specifically, in the case of abnormal occurrence during the operation of the nuclear reactor 10, if an emergency shutdown command is transferred from a control part (not shown) configured to control the operation of the nuclear reactor 10 in a process S18, the control rod drive mechanism for shutdown 32 and the control rod drive mechanism for regulating 31 are operated to simultaneously insert the regulating and shutdown rods 21 and 22 into the reactor core 12. At this time, the control rod drive mechanism for shutdown 32 and the control rod drive mechanism for regulating 31 may be operated to move the control rods for shutdown, startup, and normal operation 22, 23, and 24 and to insert them into the reactor core 12 using the driving force of the motor, the force of pressure, the force of gravity, or the restored force of the spring. However, in the process S18, when the operation of the nuclear reactor 10 is ended in a normal operating state, the control rod drive mechanism for shutdown 32 and the control rod drive mechanism for regulating 31 are respectively operated to insert the shutdown rod 22 and the control rods for startup and normal operation 23 and 24 into the reactor core 12 according to an operating signal of an operator. If the operation of the nuclear reactor 10 is ended through a process S20 or S22, the control rod drive mechanism for shutdown 32 and the control rod drive mechanism for regulating 31 is on standby until a restart of the nuclear reactor 10 in a state in which the shutdown rod 22 and the control rods for startup and normal operation 23 and 24 are inserted again into the reactor core 12, as illustrated in FIG. 10. Through the processes as described above, embodiments of the present invention may install the control rod drive mechanism at the upper or lower space of the reactor core to be disposed in the nuclear reactor, and thus may fundamentally exclude a loss-of-coolant accident or a rod ejection accident due to a damage of an external control rod drive mechanism according to the related art. Therefore, the safety of a nuclear power plant may be enhanced. Meanwhile, the embodiment has been disclosed wherein the control rod drive mechanism for regulating and the control rod drive mechanism for shutdown are respectively provided at the upper and lower sides of the reactor core. However, the present invention is not limited thereto. That is, in embodiments of the present invention, the control rod drive mechanism for shutdown may be provided at the upper side of the reactor core, and the control rod drive mechanism for regulating may be provided at the lower side of the reactor core. Alternatively, the control rod drive mechanism for shutdown and the control rod drive mechanism for regulating may be provided at the upper side of the reactor core. [Second Embodiment] FIG. 14 is a configuration view of a nuclear reactor with an in-vessel control rod drive mechanism according to a second embodiment of the present invention. As illustrated in FIG. 14, in the in-vessel control rod drive mechanism according to the second embodiment of the present invention, the control rod drive mechanism for regulating 31 may be provided at the lower side of the reactor core 12 to insert or withdraw the regulating rod 21 into/from the reactor core 12 according to an operation state of the nuclear reactor 10. The control rod drive mechanism for shutdown 32 may be provided at the upper side of the reactor core 12 to insert or withdraw the shutdown rod 22 into/from the reactor core 12 according to the operation state of the nuclear reactor 10. [Third Embodiment] FIG. 15 is a configuration view of a nuclear reactor with an in-vessel control rod drive mechanism according to a third embodiment of the present invention. As illustrated in FIG. 15, in the in-vessel control rod drive mechanism according to the third embodiment of the present invention, the control rod drive mechanism for regulating 31 may be provided at the upper side of the reactor core 12 to insert or withdraw the regulating rod 21 into/from the reactor core 12 according to the operation state of the nuclear reactor 10. The control rod drive mechanism for shutdown 32 may be provided at a higher position than the control rod drive mechanism for regulating 31 provided at the upper side of the reactor core 12 to insert or withdraw the shutdown rod 22 into/from the reactor core 12 according to the operation state of the nuclear reactor 10. [Fourth Embodiment] FIG. 16 is a configuration view of a nuclear reactor with an in-vessel control rod drive mechanism according to a fourth embodiment of the present invention. As illustrated in FIG. 16, in the in-vessel control rod drive mechanism according to the fourth embodiment of the present invention, the control rod drive mechanism for regulating 31 may be provided at the upper side of the reactor core 12 to insert or withdraw the regulating rod 21 into/from the reactor core 12 according to the operation state of the nuclear reactor 10. The control rod drive mechanism for shutdown 32 may be provided between the reactor core 12 and the control rod drive mechanism for regulating 31 provided at the upper side of the reactor core 12 to insert or withdraw the shutdown rod 22 into/from the reactor core 12 according to the operation state of the nuclear reactor 10. Meanwhile, the above-mentioned embodiments have been disclosed wherein the regulating and the shutdown rods are withdrawn to the upper side of the reactor core. However, the present invention is not limited thereto. That is, when the in-vessel control rod drive mechanism according to embodiments of the present invention is provided at the lower side of the reactor core, the embodiments may be modified such that the control rods are withdrawn to the lower side of the reactor core. [Fifth Embodiment] For example, in an in-vessel control rod drive mechanism (not shown) according to a fifth embodiment of the present invention, the control rod drive mechanism for regulating 31 may be provided at the lower side of the reactor core 12 to insert or withdraw the regulating rod 21 into/from the reactor core 12 according to the operation state of the nuclear reactor 10. The control rod drive mechanism for shutdown 32 may be provided between the reactor core 12 and the control rod drive mechanism for regulating 31 to insert or withdraw the shutdown rod 22 into/from the reactor core 12 according to the operation state of the nuclear reactor 10. [Sixth Embodiment] In an in-vessel control rod drive mechanism according to a sixth embodiment of the present invention, the control rod drive mechanism for regulating 31 may be provided at the lower side of the reactor core 12 to insert or withdraw the regulating rod 21 into/from the reactor core 12 according to the operation state of the nuclear reactor 10. The control rod drive mechanism for shutdown 32 may be provided at a lower position than the control rod drive mechanism for regulating 31 provided at the lower side of the reactor core 12 to insert or withdraw the shutdown rod 22 into/from the reactor core 12 according to the operation state of the nuclear reactor 10. As described above, in embodiments of the present invention, the control rod drive mechanism for regulating and the control rod drive mechanism for shutdown may be respectively arranged at the upper or lower side of the reactor core, or may be alternately arranged up and down at the upper or lower side of the reactor core, and thus the in-vessel control rod drive mechanism may be effectively embodied. Although a few embodiments of the present invention have been shown and described, the present invention is not limited thereto, so it should be understood that other equivalents and modifications could be made thereto without departing from the spirit and scope of the invention. Although the above-mentioned embodiments have been disclosed using the integral type reactor, embodiments of the present invention can be applied to the loop type reactor as well as the integral type reactor. Embodiments of the present invention can be applied to an in-vessel control rod drive mechanism technology in which the in-vessel control rod drive mechanism is provided at the upper or lower space of the reactor core, and the loss-of-coolant accident or the rod ejection accident due to the damage of the external control rod drive mechanism can be fundamentally excluded, and thus the safety of the nuclear power plant can be enhanced.
	claims	1. An X-ray system for imaging a target and treating thereof, comprising:a. a source providing an X-ray diverging beam having a central imaging portion directly illuminating the target and neighboring tissues and a peripheral treatment portion transformable into a converging beam;b. at least one lens transforming the peripheral treatment portion of the X-ray beam into the converging beam directed to the target;c. a first shutter located between the X-ray source and the target in the central imaging portion of the X-ray radiation; the first shutter being selectively openable such that central imaging portion of X-ray beam directly reaches the target; andd. a detector configured to detect at least a central imaging portion of the radiation directly illuminating the target after interaction thereof with the target and to provide imaging information of the target. 2. The X-ray system of claim 1, wherein the first shutter is configured to allow the imaging radiation to reach the target when the following conditions are alternatively true:a. the treatment radiation is not irradiating the target; andb. the treatment radiation is irradiating the target. 3. The X-ray system of claim 1, further comprising a filter located in the path between the X-ray source and the target, wherein the filter is configured to limit at least one of a spectral content and intensity of the imaging radiation that reaches the target. 4. The X-ray system of claim 3, wherein the filter is configured to be adjusted to allow achieving desired characteristics for images of the target acquired by the detector. 5. The X-ray system of claim 1, further comprising a second shutter being openable such that the treatment portion of X-ray beam reaches to the target. 6. The X-ray system of claim 5, wherein the second shutter is located between the X-ray source and the one or more lenses. 7. The X-ray system of claim 5, wherein the second shutter is located between the one or more lenses and the target. 8. The X-ray system of claim 5, wherein the first shutter and the second shutter are realized as an integral shutter; the first shutter and the second shutter are configured to enable at least one of the following conditions to occur:a. the target is irradiated by both the treatment radiation and the imaging radiation;b. the target is irradiated by the treatment radiation but not by the imaging radiation;c. the target is irradiated by the imaging radiation but not the treatment radiation; andd. the target is irradiated by both the treatment radiation and the imaging radiation in a time-multiplexed manner. 9. The X-ray system of claim 1, wherein at least one of the first shutter, the detector and the one or more lenses are configured to be rotated around the target, thereby enabling irradiation of the target with the treatment and/or imaging radiation from a plurality of angles. 10. The X-ray system of claim 5, wherein the detector is implemented as part of movable mechanism that allows the detector to move outside of, or within, imaging and/or treatment radiation paths. 11. The X-ray system of claim 5, wherein the first shutter is located in a direct patch between the X-ray source and the target volume. 12. The X-ray system of claim 5, wherein a third shutter is located in front of said detector. 13. The X-ray system of claim 5, wherein at least one of the shutters is actuated. 14. An X-ray system for imaging a target and treating thereof, comprising:a. a source providing an X-ray diverging beam having an imaging portion and a treatment portion angularly separated there between;b. at least one lens transforming the treatment portion of the X-ray diverging beam into a converging beam directed to a target;d. at least one reflector in an off-axis position directing the imaging portion of the X-ray diverging beam angularly to an optical axis to the target;e. a stop blocking an axially propagating portion of the X-ray diverging beam source within an internal space of at least one lens from reaching the target;f. a detector configured to detect the imaging portion of the X-ray diverging beam-after interaction thereof with the target and to provide imaging information of the target; the detector and at least one reflector are rotatable around the target such that the target is imaged at a plurality of angles.
042082471	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring now to FIG. 1, there is shown a typical thermal nuclear reactor including a sealed reactor vessel 10 housing a nuclear core 12 comprised of a plurality of fuel assemblies 14. A reactor coolant, such as one including water, enters the vessel through inlet nozzles 16, passes downward in an annular region between the vessel and a core support structure, turns and flows upward through a perforated plate 20 and through the core 12, and is discharged through outlet nozzles 22. The heat energy imparted to the coolant while passing through the core is then transferred in apparatus, not shown, ultimately for the purpose of electric power generation. The reactor coolant serves as a moderator to thermalize neutrons for the fission process. A fuel assembly 14 is shown in FIG. 3, and includes a plurality of fuel pins 24, containing nuclear fuel pellets 26, arranged in a bundle. The assembly also includes a plurality of guide thimbles 28 which provide skeletal support for the assembly and which are sized to removably receive control rods 29 of elements 30, positionable above and within the core area by means such as electromagnets 32 which act upon shafts 34 (FIG. 1) removably connected to the elements 30. A control element 30 can include one or more control rods 29. Although a "spider-type" control element is shown, including webs 35, it will be understood that many other control element and control rod configurations, including rods, bars, X-shaped and so forth, are compatible with the teachings of this invention. The control elements 30 are typically comprised of a material black to thermal neutrons, such as alloys of tantalum, silver, indium and cadmium, or boron carbide (B.sub.4 C) elements sealed within a cladding. The control elements 30 are reciprocatingly insertable into the core, between or within fuel assemblies, to control the neutron flux and hence the reactor power output. Typically the control elements 30 include those referred to as "full-length", having neutron absorbing material along their entire length, or "part length" having neutron absorbing material only along their bottom portions. The neutron flux within the core is continuously monitored by detection apparatus such as the neutron detectors 36 (FIG. 1) which are located at an elevation aligned with the elevation of the core 12. The detectors, located external to the vessel, may be fixed or laterally movable by positioning bars 38. The detectors must be operable under substantially all reactor conditions, including shutdowns for maintenance and refueling. Neutron flux detection is also accomplished on an intermittent basis by use of movable flux detectors 40 which are inserted at a predetermined rate into thimble tubes 42 which are removably positionable within the core 12 and within selected fuel assemblies 14. The thimble tubes 42, in accordance with an embodiment of this invention, are extendable above the core 12. The guide thimbles 28 of the fuel assemblies 14, in addition to receiving control rods 29, are sized to receive neutron sources 44. A typical source is shown in FIG. 4. In the prior art, the sources are, for example, shaped similar to the control rod spider element 30, but have a shorter length. The arrangement of a neutron source 44 in accordance with one embodiment of this invention is shown in FIG. 5. The source includes a fast neutron emitting material 46, sealingly encapsulated in a cladding 48 essentially black to thermal neutrons. The preferred source material 46 is a combination of plutonium-238 and beryllium, and the preferred cladding 48 includes cadmium, such as a combination of silver, indium and cadmium. The plutonium-238, which has a half-life of approximately 89 years, emits alpha particles. The alpha particles strike the beryllium, and result in (.alpha., n) reactions, emitting fast neutrons. The fast neutrons can readily pass through the cladding and may be thermalized in the surrounding medium or can continue through the reactor vessel to activate the neutron detectors 36. Cadmium is preferred as the source cladding 48 as it significantly reduces neutron-induced heat generation within the source 44. The heat generation results primarily from fission of plutonium-238 nuclei by thermal and resonance energy neutrons that penetrate the cladding 48 during power operation. Preliminary calculations indicate that the heat generation rate due to fissioning of plutonium-238 by low energy neutrons in, for example, a 100 curie source will be reduced to an acceptable maximum of 20 watts by encapsulation in cladding 48 having a wall thickness as shown in Table I. Pure cadmium is listed for comparison purposes only, as the melting point of cadmium, 321.degree. C. (610.degree. F.), limits its structural integrity within a reactor environment. The relative percentages of the compounds are given by weight, and although adding up to 100%, do not preclude some impurities. TABLE I ______________________________________ Required Cladding Material Thickness, Inches ______________________________________ Cadmium 0.008 65% Silver, 35% Cadmium 0.023 80% Silver, 15% Indium, 5% Cadmium 0.075 ______________________________________ The source 44 also preferably includes a plenum 50 within the sealed cladding 48, in flow communication with the source material 46. The plenum 50, or void space, serves to collect gas within the source, most particularly helium resulting from the emitted alpha particles. The cladding 48 can be sealed in a variety of manners, including the use of welded end plugs 52. The described neutron source will have an extended life relative to prior art sources as a result of the long half-life of plutonium-238 and the protection afforded the plutonium by the cadmium cladding. However, the life of the described source, or prior art sources, can be further extended if the source does not reside in the core region 12 continuously throughout a fuel cycle. Another embodiment is an arrangement which provides an extended life, and is shown in FIG. 5. The neutron source 44, or any source, is here affixed to a control rod 29 or control element 30, which is slidingly movable within the core region. The rod 29 can specifically be received within one of the tubular guide thimbles 28 of a fuel assembly 14. Preferably the source 44 is sealed within a control rod cladding 54, which also seals a neutron absorbing control material 56 as well known in the art. As the control rods or elements are reciprocatingly inserted and withdrawn from the reactor core 12 in order to control reactor power and shutdown reactivity, a neutron source affixed to an element insertable at shutdown conditions and withdrawn at power operation will extend the source operating life. Additionally, as indicated by FIG. 6, a singular rod 29 can be separately fabricated and field assembled to a complete control element 30. The rod 29 can, for example, be field welded to the web 35 or a web receptacle 60. It will be apparent that the extended life benefit provided by the incorporation of a source with a control means is applicable to other source types including those of the prior art, in addition to the preferred source disclosed. Dependent upon the strength of the source and the configuration of the core and the surrounding components, the source 44 is radially positioned so as to activate the detectors 36 at low power. Typically a source can be included in one or more of the outermost rods 58 of a radially outermost control element 30, as shown in FIG. 2. The control elements utilized are preferably within the boundary of the outer three rows of fuel assemblies within the exemplary core. It will also be apparent that the control elements 30 including neutron sources, for example, two in a reactor core, can advantageously be interconnected to the balance of the control elements and reactor control systems to achieve desired operational movements. For example, when the reactor is subcritical, the source-bearing control elements can be electrically disconnected from the normal control banks or groups with which they are associated to avoid potential inadvertent movement of the sources and disruption of neutron flux level monitoring. During an approach to criticality, the source-bearing elements would remain disengaged, and within the core, while shutdown and first control groups are withdrawn from the core to establish criticality. The flux level accordingly increases to well above the minimum source level range during this operation. At this point, the source bearing elements should then be withdrawn to form a normal operating configuration, and electrically connected to the normal control means. Alternatively, a separate source shutdown control element group, including the source-bearing elements, can be defined within the normal control element programming control system. Additional uses of the inventive teachings can easily be visualized. For example, the source 44 can be used for calibration of the movable flux detectors 40. This can be accomplished by orienting a thimble tube 42 proximate a control element 30 including the neutron source. The thimble tube 42 would be inserted so that its extremity rises above the top of the core, and the movable detector inserted to also rise above the core. The neutron source can then be laterally aligned with the movable detector 40, and used as a source of calibration for the detector 40. It will be apparent that utilization of the inventive teachings provides distinct advantages over prior art neutron sources. The invention permits reduction of secondary source use with an associated reduction in operating costs and core complexity. It can eliminate the need for special orificing which has been required when fixed neutron sources are placed in fuel assemblies. A combination neutron source and control element also markedly reduces the neutron and gamma radiation exposure of the source and improves its long term mechanical integrity. Refueling procedures are also simplified by reducing the number of components which must be handled. And, in addition to extending the life of neutron emitting sources, the teachings can also be utilized for calibration of movable in-core flux detectors. It will be further apparent that additional modifications and benefits are possible in view of the above teachings. It therefore is to be understood that within the scope of the appended claims, the invention may be practiced other than as specifically described.
	summary
061608629	claims	1. A primary target for the production of fission products comprising: a.) an inner cylinder having an outer surface, a first end, a second end, a first raised shoulder at the first end and a second raised shoulder at the second end; b.) a foil of fissionable material having a first surface and a second surface, whereby the foil forms a layer on the outer surface of the inner cylinder so that the foil first surface contacts the outer surface of the inner cylinder; c.) an outer barrier having a first surface and a second surface, whereby the outer barrier forms a layer on the foil such that the outer barrier first surface contacts the foil second surface; d.) an outer hollow cylinder selected from a material having a coefficient of thermal expansion lower than that of inner cylinder and adapted to receive the inner cylinder, having an inner surface, a first end, and a second end, so that the inner surface of the outer hollow cylinder forms a layer on the outer hollow barrier such that the outer cylinder inner surface contacts the outer barrier second surface, the outer hollow cylinder first end contacts the inner cylinder first raised shoulder and the outer hollow cylinder second end contacts the inner cylinder second raised shoulder, wherein the outer surface of the inner cylinder is tapered relative to the inner surface of the outer cylinder to ensure good physical contact between contact surfaces of the inner cylinder, the foil, the outer barrier and the outer cylinder; and e.) a means for attaching the outer cylinder first end to the inner cylinder first raised shoulder and the outer cylinder second end to the inner cylinder second raised shoulder, whereby the foil and the outer barrier are contained between the inner and outer cylinders. a.) an inner cylinder having an outer surface, a first end, a second end, a first raised shoulder at the first end and a second raised shoulder at the second end; b.) an inner barrier having a first surface and a second surface, whereby the inner barrier second surface contacts the inner cylinder outer surface; c.) a foil of fissionable material comprised of low enriched uranium or plutonium metal having a first surface and a second surface, whereby the inner barrier forms a layer on the foil first surface so that the inner barrier first surface contacts the foil first surface; d.) an outer barrier compromised of the same material as the inner barrier having a first surface and a second surface, whereby the outer barrier forms a layer on the foil such that the outer barrier first surface contacts the foil second surface; e.) an outer hollow cylinder selected from a material having a coefficient of thermal expansion lower than that of inner cylinder and adapted to receive the inner cylinder, having an inner surface, a first end, and a second end, so that the inner surface of the outer hollow cylinder forms a layer on the outer barrier such that the outer cylinder inner surface contacts the outer barrier second surface, the outer cylinder first end contacts the inner cylinder first raised shoulder and the outer cylinder second end contacts the inner cylinder second raised shoulder, wherein the outer surface of the inner cylinder is tapered relative to the inner surface of the outer cylinder to ensure good physical contact and thermal conductivity between contact surfaces of the inner cylinder, the inner barrier, the foil, the outer barrier and the outer cylinder; and f) a first weld attaching the outer cylinder first end to the inner surface first raised shoulder and a second weld attaching the outer cylinder second end to the inner cylinder second raised shoulder, whereby the foil and the inner and outer barriers are contained between the inner and outer cylinders. 2. A primary target as recited in claim 1 wherein the outer barrier is a metal selected from the group consisting of copper, nickel, iron and zinc. 3. A primary target as recited in claim 1 including an inner barrier having a first surface and a second surface, whereby the inner barrier forms a layer on the foil such that the inner barrier first surface contacts the foil first surface and the inner barrier second surface contacts the inner cylinder outer surface. 4. A primary target as recited in claim 3 wherein the inner barrier and the outer barrier are metals selected from the group consisting of copper, nickel, iron and zinc. 5. A primary target as recited in claim 4 wherein the inner barrier and the outer barrier are comprised of the same materials. 6. A primary target as recited in claim 4 wherein the inner cylinder and the outer cylinder are nonfissionable material selected from the group consisting of stainless steel, nickel, nickel alloys, zirconium, zircaloy, aluminum or zinc coated aluminum. 7. A primary target as recited as recited in claim 6 wherein the inner cylinder and outer cylinder are comprised of different materials. 8. A primary target as recited in claim 7 wherein the inner cylinder is comprised of aluminum and the outer cylinder is comprised of zirconium. 9. A primary target as recited in claim 7 wherein the inner cylinder is comprised of stainless steel and the outer cylinder is comprised of zirconium. 10. The invention as recited in claim 7 wherein the foil of fissionable material is low enriched uranium or plutonium metal. 11. A primary target as recited in claim 10 wherein the inner cylinder first and second raised shoulders are raised to the same height relative to the outer surface of the inner cylinder to accommodate a predetermined thickness of the foil. 12. A primary target as recited in claim 11 wherein the predetermined thickness of the foil is selected from a range of between approximately 0.001 inches to 0.01 inches. 13. A primary target as recited in claim 10 wherein the means for attaching is a weld. 14. A primary target as recited in claim 10 wherein the inner cylinder and the outer cylinder have a thickness selected from a range of between approximately 0.025 inches and 0.060 inches. 15. A primary target for the production of fission products comprising: 16. A primary target as recited in claim 15 wherein the inner barrier and the outer barrier are metals selected from the group consisting of copper, nickel, iron and zinc. 17. A primary target as recited in claim 16 wherein the inner cylinder and the outer cylinder are nonfissionable material selected from the group consisting of stainless steel, nickel, nickel alloys, zirconium, zircaloy, aluminum or zinc coated aluminum. 18. A primary target as recited as recited in claim 17 wherein the inner cylinder and outer cylinder are comprised of different materials. 19. A primary target as recited in claim 18 wherein the inner cylinder first and second raised shoulders are raised to the same height relative to the outer surface of the inner cylinder to accommodate a predetermined thickness of the foil. 20. A primary target as recited in claim 19 wherein the inner cylinder and the outer cylinder have a thickness selected from a range of between approximately 0.025 inches and 0.060 inches.
	claims	1. A radiation source comprising:a first electrode having an aperture substantially centered around a central axis of said radiation source for passing to pass electromagnetic radiation from said radiation source; and a second electrode spaced apart from said first electrode to form a gap therebetween, said gap defining a discharge region, said gap being supplied, in use of the radiation source, with a working vapor, wherein one of said first electrode and or said second electrode includes a hollow cavity in communication with said discharge region through an aperture that has a substantially annular configuration around the central axis of said radiation source, said hollow cavity is being supplied, in use of the radiation source, with a driver gas. 2. A radiation source according to claim 1, wherein said first electrode is an anode and said second electrode is a cathode. 3. A radiation source according to claim 1, wherein said first electrode is a cathode and said second electrode is an anode. 4. A radiation source according to claim 1, wherein said cavity has a substantially annular configuration around the central axis of the radiation source. 5. A radiation source according to claim 1, wherein said first electrode and said second electrode are connected to different electrical potentials such that, in use of the radiation source, a plasma discharge is initiated in said driver gas inside said hollow cavity, followed by a compression of the plasma of the driver gas towards said central axis of said radiation source in which said plasma of the driver gas encounters the working vapor to create a plasma in the working vapor, the plasma in the working vapor emitting said electromagnetic radiation. 6. A radiation source according to claim 1, wherein said driver gas comprises at least one selected from the group comprising helium, neon, argon and hydrogen. 7. A radiation source according to claim 1, further comprising:a shutter disposed in the vicinity of the aperture of the first electrode, wherein said shutter is adapted to substantially block particles formed in said discharge region. 8. A radiation source according to claim 7, wherein said shutter includes a flywheel. 9. A radiation source according to claim 7, wherein said shutter is adapted to open the aperture of the first electrode to let radiation pass through and to close the aperture of the first electrode to substantially block said particles. 10. A radiation source according to claim 1, wherein said working vapor is supplied, in use of the radiation source, in a region proximate said central axis of said gap between said first electrode and said second electrode. 11. A radiation source according to claim 10, wherein said working vapor is supplied, in use of the radiation source, along said central axis. 12. A radiation source according to claim 1, wherein said working vapor comprises xenon. 13. A radiation source according to claim 1, wherein said working vapor comprises at least one selected from the group comprising lithium vapor and tin vapor. 14. A radiation source according to claim 13, further comprising:a reservoir adapted to contain a material comprising at least one of selected from the group comprising lithium and tin; a heater arranged to heat at least a portion of said reservoir so as to create a vapor from said material; and a fluid passageway in communication with said reservoir to allow said vapor to enter said gap between said first electrode and said second electrode. 15. A radiation source according to claim 13, further comprising:a reservoir adapted to contain a material comprising at least one of selected from the group comprising lithium and tin; a first heater arranged to heat at least a portion of said reservoir so as to create a liquid from said material; and a fluid passageway in communication with said reservoir, said liquid created from said material is drawn and configured to draw inside said fluid passageway by capillary action said liquid created from said material. 16. A radiation source according to claim 15, further comprising:a second heater in contact with a portion of said fluid passageway, wherein said second heater is arranged to heat at least a portion of said liquid drawn inside said fluid passageway so as to create a vapor from said liquid and to allow said vapor to enter said gap between said first electrode and said second electrode by capillary action. 17. A radiation source according to claim 15, wherein said fluid passageway includes a tubular section. 18. A radiation source according to claim 15, wherein said fluid passageway includes a porous rod. 19. A radiation source according to claim 18, wherein said porous rod is terminated at one of its ends with a chamber, said chamber being adapted to collect the at least one of the lithium vapor and the tin vapor. 20. A radiation source according to claim 19, wherein said chamber includes an opening through which configured to allow the at least one of said lithium vapor and tin vapor escapes to escape to said gap between said first electrode and said second electrode. 21. A radiation source according to claim 1, further comprising:an electrical insulator disposed between said first electrode and said second electrode; and a canal leading to a said gap between said first electrode and said second electrode, wherein said electrical insulator is disposed inside said canal away from, in use of the radiation source, said working vapor. 22. A radiation source according to claim 21,wherein said canal defines a path along which, in use of the radiation source, said working vapor condenses to form a liquid material. 23. A radiation source according to claim 22, wherein, in use of the radiation source, a temperature along said path is less than or equal to 300° C. 24. A radiation source according to claim 21 22, wherein said canal is inclined relative to a wall of a reservoir in said radiation source such that, in use of the radiation source, the liquid material is collected by gravity in said reservoir. 25. A radiation source according to claim 21, wherein said canal comprises a curved portion. 26. A radiation source according to claim 1, further comprising a trigger electrode, wherein at least a portion of said electrode is disposed within said hollow cavity. 27. A radiation source according to claim 26, further comprising an electrical circuit constructed and arranged to apply a voltage pulse to said trigger electrode. 28. A radiation source according to claim 27, wherein said electrical circuit comprises a transformer having primary and secondary windings, said primary windings being in electrical communication with a voltage source to supply said voltage pulse and said secondary windings being in electrical communication with one of said first electrode and or second electrode and said trigger electrode. 29. A radiation source according to claim 1, wherein said radiation source is adapted to generate a beam of radiation having a wavelength between about 5 nm and about 20 nm. 30. A lithographic projection apparatus comprising:a radiation system adapted to provide a projection beam of radiation; a support structure adapted to support a patterning structure to pattern the projection beam according to a desired pattern; a substrate table adapted to hold a substrate; and a projection system disposed between said support structure and said substrate table, said projection system being configured to project the patterned beam onto a target portion of the substrate, wherein said radiation system comprises: a first electrode having an aperture centered around a central axis of said radiation source system; and a second electrode spaced apart from said first electrode to form a gap therebetween, said gap defining a discharge region, said gap being supplied, in use of the radiation system, with a working vapor, wherein one of said first electrode and or said second electrode includes a hollow cavity in communication with said discharge region through an aperture that has a substantially annular configuration around the central axis of said radiation source system, said hollow cavity is being supplied, in use of the radiation system, with a driver gas. 31. A radiation source comprising:a plasma chamber adapted to house a plasma, said plasma chamber having an aperture through which a radiation emitted by said plasma passes is configured to pass; a shutter disposed in a vicinity of said aperture, wherein said shutter is adapted to substantially block particles formed in said plasma. 32. A radiation source according to claim 31,wherein said shutter includes a flywheel. 33. A radiation source according to claim 31,wherein said shutter is adapted to open the aperture to let said radiation emitted by said plasma to pass through and to close the aperture to substantially block said particles formed in said plasma. 34. A radiation source according to claim 31,wherein a first wall of said chamber forms a first electrode and a second wall of said chamber forms a second electrode. 35. A radiation source comprising:a source of material including at least one of selected from the group comprising lithium and tin; an electrode disposed in the vicinity of said source, wherein said electrode is configured to induce formation of a plasma in said at least one of the lithium and the tin, said plasma emitting a radiation having a wavelength in an extreme ultraviolet range of wavelengths. 36. A radiation source according to claim 35, wherein said extreme range of wavelengths is between about 5 nm and about 20 nm. 37. A radiation source according to claim 35, further comprising:a heater arranged to heat at least a portion of said source of material so as to create a vapor from said material; and a fluid passageway in communication with said source of material to allow said vapor from said material to enter a region in said radiation source in which said plasma takes place during use of the radiation source. 38. A radiation source according to claim 37, wherein said fluid passageway includes a porous rod. 39. A radiation source according to claim 38, wherein said porous rod is terminated at one of its ends with a chamber, said chamber being adapted to collect a vapor from said material. 40. A radiation source according to claim 39, wherein said chamber includes an opening through which said vapor from said material escapes, in use of the radiation source, to said region in said radiation source in which said plasma takes place during use of the radiation source. 41. A radiation source comprising a first electrode and a second electrode that are configured and arranged to create a discharge in a gas in a discharge region defined between the first and second electrodes and to form a plasma from a working vapor so as to generate extreme ultraviolet radiation, wherein the second electrode further at least partly defines a hollow cavity in communication with the discharge region through an aperture that has a substantially annular configuration around a central axis of the radiation source. 42. A radiation source according to claim 41, wherein the first electrode is an anode and the second electrode is a cathode. 43. A radiation source according to claim 41, wherein the first electrode is a cathode and the second electrode is an anode. 44. A radiation source according to claim 41, wherein the hollow cavity has a substantially annular configuration around the central axis of the radiation source. 45. A radiation source according to claim 41, wherein, in use of the radiation source, a driver gas is supplied to the hollow cavity. 46. A radiation source according to claim 45, wherein the driver gas comprises at least one selected from the group comprising helium (He), neon (Ne), argon (Ar) and hydrogen (H2). 47. A radiation source according to claim 41, wherein the working vapor, in use of the radiation source, is supplied in a region proximate the central axis of the space between the first and second electrodes. 48. A radiation source according to claim 47, wherein the working vapor, in use of the radiation source, is supplied along the central axis. 49. A radiation source according to claim 41, wherein the working vapor comprises xenon. 50. A radiation source according to claim 41, wherein the working vapor comprises at least one selected from the group comprising lithium vapor and tin vapor. 51. A radiation source according to claim 50, wherein the radiation source comprises a reservoir to contain a material comprising at least one selected from the group comprising lithium and tin, a heater arranged to heat the reservoir so as to create vapor from the material, and a fluid communication to allow the vapor to enter the space between the first and second electrodes. 52. A radiation source according to claim 50, wherein the radiation source comprises a reservoir to contain a material comprising at least one selected from the group comprising lithium and tin, a heater arranged to heat the reservoir so as to create a liquid from the material, and a fluid communication to allow the liquid to enter the space between the first and second electrodes by capillary action. 53. A radiation source according to claim 50, further comprising an electrical insulator between the first and second electrodes, and a path between a region of the space between the first and second electrodes where the working vapor is provided in use of the radiation source and the electrical insulator is constructed and arranged to define a space for the working vapor to condense along the path. 54. A radiation source according to claim 41, further comprising a trigger electrode at least partially disposed within the hollow cavity. 55. A radiation source according to claim 54, further comprising an electrical circuit constructed and arranged to apply a voltage pulse to the trigger electrode. 56. A radiation source according to claim 55, wherein the electrical circuit comprises a transformer having primary and secondary windings, the primary windings being in electrical communication with a voltage source to supply the voltage pulse and the secondary windings being in electrical communication with the first electrode or second electrode and the trigger electrode. 57. A radiation source according to claim 41, adapted to generate a beam of extreme ultraviolet radiation having a wavelength of between about 5 nm and about 20 nm. 58. A radiation source according to claim 57, wherein the beam of extreme ultraviolet radiation has a wavelength between about 9 nm and about 16 nm. 59. A radiation source according to claim 41, wherein, in use of the radiation source, a driver gas is supplied to the hollow cavity and the first electrode and the second electrode are connected to different electrical potentials such that, in use of the radiation source, a plasma discharge is initiated in the driver gas inside the hollow cavity, followed by a compression of the plasma of the driver gas towards the central axis of the radiation source in which the plasma of the driver gas encounters the working vapor to create a plasma in the working vapor, the plasma in the working vapor emitting the extreme ultraviolet radiation. 60. A lithographic projection apparatus comprising:a radiation system to provide a beam of radiation, the radiation system comprising a first electrode and a second electrode that are configured and arranged to create a discharge in a gas in a discharge region defined between the first and second electrodes and to form a plasma from a working vapor so as to generate extreme ultraviolet radiation, wherein the second electrode further at least partly defines a hollow cavity in communication with the discharge region through an aperture that has a substantially annular configuration around a central axis of the radiation system; a support structure to support a patterning structure, the patterning arranged to pattern the beam according to a desired pattern; a substrate table to hold a substrate; and a projection system to project the beam as patterned onto a target portion of the substrate. 61. A lithographic projection apparatus according to claim 60, wherein, in use of the radiation system, a driver gas is supplied to the hollow cavity. 62. A lithographic projection apparatus according to claim 61, wherein the first electrode and the second electrode are connected to different electrical potentials such that, in use of the radiation system, a plasma discharge is initiated in the driver gas inside the hollow cavity, followed by a compression of the plasma of the driver gas towards the central axis of the radiation system in which the plasma of the driver gas encounters the working vapor to create a plasma in the working vapor, the plasma in the working vapor emitting the extreme ultraviolet radiation. 63. A lithographic device manufacturing method comprising:providing a beam of extreme ultraviolet radiation using a radiation source comprising a first electrode and a second electrode that are configured and arranged to create a discharge in a gas in a discharge region defined between the first and second electrodes and to form a plasma from a working vapor so as to generate the extreme ultraviolet radiation, wherein the second electrode further at least partly defines a hollow cavity in communication with the discharge region through an aperture that has a substantially annular configuration around a central axis of the radiation source; patterning the beam of radiation to produce a patterned beam of radiation; and projecting the patterned beam of radiation onto a target portion of a layer of radiation-sensitive material on a substrate. 64. A lithographic device manufacturing method according to claim 63, further comprising supplying a driver gas to the hollow cavity. 65. A lithographic device manufacturing method according to claim 64, comprising apply different potentials to the first electrode and the second electrode such that a plasma discharge is initiated in the driver gas inside the hollow cavity, followed by a compression of the plasma of the driver gas towards the central axis of the radiation source in which the plasma of the driver gas encounters the working vapor to create a plasma in the working vapor, the plasma in the working vapor emitting the extreme ultraviolet radiation. 66. A lithographic device manufacturing method according to claim 64, wherein the driver gas comprises at least one selected from the group comprising helium (He), neon (Ne), argon (Ar) and hydrogen (H2). 67. A lithographic device manufacturing method according to claim 63, wherein the working vapor comprises at least one selected from the group comprising lithium vapor, tin vapor and xenon.
	abstract	The present invention relates to an X-ray imaging apparatus including an X-ray source, a grating that divides diverging X-rays irradiated from the X-ray source, and a detector that detects X-rays which are divided by the grating and pass through a sample. The grating includes a plurality of transparent objects which pass the diverging X-rays and a plurality of opaque objects that shield the diverging X-rays. A focused position at which a plurality of extended lines intersect each other and the X-ray source are arranged in different position. The extended lines are formed by extending center lines which connect a center of the X-ray source side of each of the plurality of opaque objects facing the X-ray source with a center of the detector side of each of the plurality of opaque objects facing the detector.
050769944	summary	TECHNICAL FIELD The present invention relates generally to control rod drives used in nuclear reactors and, more specifically, to a lock assembly effective for preventing rotation of the control rod drive when actuated. BACKGROUND ART In one type of nuclear reactor, control rods are selectively inserted and withdrawn from a nuclear reactor vessel for controlling the operation thereof. Each of the control rods is typically positioned by a conventional control rod drive which includes a ball screw or spindle threadingly engaging a ball nut for raising and lowering the ball nut as the spindle is rotated either clockwise or counterclockwise. A hollow piston rests upon the ball nut at one end thereof and at its other end is conventionally joined to the control rod. Displacement of the ball nut provides displacement of the piston which in turn inserts or withdraws the control rod in the reactor vessel. In order to achieve faster insertion of the control rod than could be obtained by normal rotation of the ball spindle, which is conventionally referred to as a scram operation, a rapid flow of high-pressure water is injected through the control rod drive past the piston for lifting the piston off the ball nut in a relatively short time for quickly inserting the control rod into the reactor vessel. The high-pressure water is channeled to the control rod drive through a scram line pipe attached to a high-pressure water accumulator. In one type of occurrence which allows for rapid backflow of the water past the piston, due to, for example, a break in the scram line, the backflow may cause a large reverse pressure on the piston which in turn provides a back force on the control rod ball nut. This back force can cause reverse rotation of the ball spindle with corresponding withdrawal of the control rod. Withdrawal of one of the control rods due to such a backflow occurrence may cause damage to adjacent fuel in the reactor vessel, requiring replacement thereof leading to undesirable down time of the reactor and economic losses. In order to prevent the above occurrence, a conventional electromechanical brake is provided in the control rod drive for holding the ball spindle from rotating unless the brake is energized. The brake is sized for restraining rotation of the ball spindle against such forces due to backflow of water over the piston when the control rod drive motor is not operating. And, when the control rod drive motor is operating, the motor itself is sized for providing adequate torque for resisting the forces due to the backflow of water in the event of the above-described occurrence. To ensure operability of the brake, the brake is periodically tested. However, the brakes are located adjacent to the reactor vessel, which is inaccessible during operation of the reactor due to the radiation field emanating from the reactor vessel. The radiation field continues at reduced levels also during shutdown of the reactor, which would require inspectors to wear suitable protective clothing and limit their time in the area. In one nuclear reactor embodiment, there are about 205 control rod drives, including a respective number of brakes, which would necessarily require a substantial amount of time for testing all of the brakes. Testing of the brakes during reactor shutdown would, therefore, be relatively costly to accomplish, which is additionally economically undesirable since the reactor is not operating for producing power. Since conventional electromechanical brakes typically utilize braking pads for restraining rotation of a rotor disc, they are subject to slippage. Slippage can result in undesirable partial withdrawal of the control rod during backflow occurrence, and also requires additional means for effectively testing the torque-resisting capability of the brake. OBJECTS OF THE INVENTION Accordingly, one object of the present invention is to provide a new and improved lock assembly for preventing rotation of a shaft. Another object of the present invention is to provide a lock assembly effective for providing a positive rotational restraint of the shaft in one direction while allowing rotation in an opposite direction. Another object of the present invention is to provide a relatively simple and compact lock assembly for a shaft. Another object of the present invention is to provide a lock assembly for preventing rotation of a control rod drive for a nuclear reactor and which may be actuated and tested remotely therefrom. DISCLOSURE OF INVENTION A lock assembly is disclosed for selectively preventing rotation of a shaft, such as a shaft used in a control rod drive for a nuclear reactor. The lock assembly includes a stationary housing for receiving the shaft, and a gear fixedly joined to the shaft. The gear includes a plurality of circumferentially spaced gear teeth. An elongate key is slidably joined to the housing and has at least one locking tooth facing the gear teeth. Means are provided for selectively positioning the key in an engaged position wherein the locking tooth and gear teeth prevent rotation of the shaft in a first direction, and in a disengaged position for allowing the shaft to rotate without obstruction between the gear teeth and the locking tooth.
	abstract	Aberration-corrected charged-particle optical apparatus improving the resolution of charged-particle optical systems by eliminating or minimizing optical aberrations. The apparatus comprises a source of charged particles and a plurality of charged-particle lenses including non-round lenses, energized in such manner so as to correct axial aberrations of orders up to and including fifth order. The non-round lenses comprise quadrupoles and octupoles disposed in such manner that fifth order combination aberrations are precisely controlled in addition to third order aberrations. The resultant apparatus very significantly improves on resolution attainable with non-aberration corrected charged-particle round lenses. It also improves on resolution attainable with correctors of third order aberrations only.
052689514	abstract	A x-ray scanning method involves the stops of directing an x-ray beam at a collimating first mirror having the capability of altering the source to mirror location and/or grazing angle of incidence. The beam is then reflected from a flat second mirror capable of a scanning motion by linear translation with an optional accompanying angular change in the grazing angle of incidence of the beam on the second mirror.
	description	This application is a continuation of U.S. patent application Ser. No. 15/603,222, filed on May 23, 2017, which claims priority to U.S. Provisional Patent Application No. 62/342,028, filed on May 26, 2016, the entire contents of each of which are fully incorporated herein by reference. Contained herein is material that is subject to copyright protection. The copyright owner has no objection to the facsimile reproduction by anyone of the patent document or the patent disclosure, as it appears in the United States Patent and Trademark Office patent file or records, but otherwise reserves all rights to the copyright whatsoever. The following notice applies to the software, screenshots and data as described below and in the drawings hereto and All Rights Reserved. This disclosure relates generally to modular shielding for storage containers, particularly for storage containers comprising substances that either emit unwanted elements, compounds, or materials to the environment, or require protection from the environment. Certain elements, compounds, or materials radiate unwanted or harmful components when stored. One example of this type of material is nuclear waste. Nuclear waste currently in storage comes from three principal sources: spent fuel from commercial or research reactors, liquid waste from the reprocessing of spent fuel, and waste from the nuclear weapons and propulsions industry. Most of the storage concerns relate to so-called ‘intermediate and high level’ nuclear waste components, which are highly radioactive, often requiring cooling and containment because their decay gives off heat and radiation, and have an extremely long half-life. Long-term storage of radioactive waste is aided by the stabilization of the waste into a form which will neither react nor degrade for extended periods of time. Currently, vitrification is an accepted practice to achieve this stabilization. The vitrification process requires nuclear waste to be mixed with glass forming media (soil or zeolite, as an example), and heated to the point that the mixture melts. Once cooled, the result is that the nuclear waste is effectively entrained in glass, with reduced chances of leakage and exposure to the environment. Some vitrification methods allow the vitrification process to occur in the actual storage container, thereby minimizing waste handling and reducing contamination possibilities from processing. This type of vitrification is known as in-container vitrification, or ICV™. The containers used for this process are called ICV™ Containers. Once processed through vitrification, the ICV™ containers are stored, either temporarily or long term. Shielding is used to mitigate potential harmful energy from the radioactive decay of certain elements. Within current shielding for ICV™ storage systems there is little room for reconfiguration and adjustability of the shielding. Additionally, with current systems more shielding is being used than is necessary which is not economical both from materials and storage capacity standpoints. The converse can be true, i.e. some stored compounds or materials need shielding from the environment around them. What is needed is an adjustable, compact, modular shielding system for short or long-term storage containers requiring shielding to prevent either the escape of the contents, particles, or rays, or prevent the ingress of particles or rays to the container. So as to reduce the complexity and length of the Detailed Specification, Applicant(s) herein expressly incorporate(s) by reference all of the following materials identified in each paragraph below. The incorporated materials are not necessarily “prior art” and Applicant(s) expressly reserve(s) the right to swear behind any of the incorporated materials. System for Vitrification Container with Removable Shield Panels, Ser. No. 62/342,028, filed May 26, 2016, which is herein incorporated by reference in its entirety, and to which this application claims priority. System and Method for a Robotic Manipulator Arm, Ser. No. 15/591,978 filed May 10, 2017, with a priority date of May 16, 2016, which is hereby incorporated by reference in its entirety. Mobile Processing System, Ser. No. 14/748,535, filed Jun. 24, 2015, with a priority date of Jun. 24, 2014, which is herein incorporated by reference in its entirety. Ion Specific Media Removal from Vessel for Vitrification, Ser. No. 15/012,101 filed Feb. 1, 2016, with a priority date of Feb. 1, 2015, which is hereby incorporated by reference in its entirety. System and Method for an Electrode Seal Assembly, Ser. No. 15/388,299 filed Dec. 22, 2016, with a priority date of Dec. 29, 2015, which is herein incorporated by reference in its entirety. Methods for Melting of Materials to be Treated, U.S. Pat. No. 7,211,038 filed Mar. 25, 2001, with a priority date of Sep. 25, 2001, which is herein incorporated by reference in its entirety. Methods for Melting of Materials to be Treated, U.S. Pat. No. 7,429,239 filed Apr. 27, 2007, with a priority date of Sep. 25, 2001, which is herein incorporated by reference in its entirety. Vitrification of Waste with Continuous Filling and Sequential Melting, U.S. Pat. No. 6,283,908 filed May 4, 2000, with a priority date of May 4, 2000, which is herein incorporated by reference in its entirety. Applicant(s) believe(s) that the material incorporated above is “non-essential” in accordance with 37 CFR 1.57, because it is referred to for purposes of indicating the background or illustrating the state of the art. However, if the Examiner believes that any of the above-incorporated material constitutes “essential material” within the meaning of 37 CFR 1.57(c)(1)-(3), applicant(s) will amend the specification to expressly recite the essential material that is incorporated by reference as allowed by the applicable rules. Aspects and applications presented here are described below in the drawings and detailed description. Unless specifically noted, it is intended that the words and phrases in the specification and the claims be given their plain, ordinary, and accustomed meaning to those of ordinary skill in the applicable arts. The inventors are fully aware that they can be their own lexicographers if desired. The inventors expressly elect, as their own lexicographers, to use only the plain and ordinary meaning of terms in the specification and claims unless they clearly state otherwise and then further, expressly set forth the “special” definition of that term and explain how it differs from the plain and ordinary meaning. Absent such clear statements of intent to apply a “special” definition, it is the inventors' intent and desire that the simple, plain and ordinary meaning to the terms be applied to the interpretation of the specification and claims. The inventors are also aware of the normal precepts of English grammar. Thus, if a noun, term, or phrase is intended to be further characterized, specified, or narrowed in some way, then such noun, term, or phrase will expressly include additional adjectives, descriptive terms, or other modifiers in accordance with the normal precepts of English grammar. Absent the use of such adjectives, descriptive terms, or modifiers, it is the intent that such nouns, terms, or phrases be given their plain, and ordinary English meaning to those skilled in the applicable arts as set forth above. Further, the inventors are fully informed of the standards and application of the special provisions of 35 U.S.C. § 112, 16. Thus, the use of the words “function,” “means” or “step” in the Detailed Description or Description of the Drawings or claims is not intended to somehow indicate a desire to invoke the special provisions of 35 U.S.C. § 112, 16, to define the systems, methods, processes, and/or apparatuses disclosed herein. To the contrary, if the provisions of 35 U.S.C. § 112, 16 are sought to be invoked to define the embodiments, the claims will specifically and expressly state the exact phrases “means for” or “step for, and will also recite the word “function” (i.e., will state “means for performing the function of . . . ”), without also reciting in such phrases any structure, material or act in support of the function. Thus, even when the claims recite a “means for performing the function of . . . ” or “step for performing the function of . . . ”, if the claims also recite any structure, material or acts in support of that means or step, or that perform the recited function, then it is the clear intention of the inventors not to invoke the provisions of 35 U.S.C. § 112, 16. Moreover, even if the provisions of 35 U.S.C. § 112, 6 are invoked to define the claimed embodiments, it is intended that the embodiments not be limited only to the specific structure, material or acts that are described in the preferred embodiments, but in addition, include any and all structures, materials or acts that perform the claimed function as described in alternative embodiments or forms, or that are well known present or later-developed, equivalent structures, material or acts for performing the claimed function. Elements and acts in the figures are illustrated for simplicity and have not necessarily been rendered according to any particular sequence or embodiment. In the following description, and for the purposes of explanation, numerous specific details, process durations, and/or specific formula values are set forth in order to provide a thorough understanding of the various aspects of exemplary embodiments. However, it will be understood by those skilled in the relevant arts, that the apparatus, systems, and methods herein may be practiced without these specific details, process durations, and/or specific formula values. It is to be understood that other embodiments may be utilized and structural and functional changes may be made without departing from the scope of the apparatus, systems, and methods herein. In other instances, known structures and devices are shown or discussed more generally in order to avoid obscuring the exemplary embodiments. In many cases, a description of the operation is sufficient to enable one to implement the various forms, particularly when the operation is to be implemented in software. It should be noted that there are many different and alternative configurations, devices, and technologies to which the disclosed embodiments may be applied. The full scope of the embodiments is not limited to the examples that are described below. In the following examples of the illustrated embodiments, references are made to the accompanying drawings which form a part hereof, and in which is shown by way of illustration various embodiments in which the systems, methods, processes, and/or apparatuses disclosed herein may be practiced. It is to be understood that other embodiments may be utilized and structural and functional changes may be made without departing from the scope. A removable shield panel (RSP) system is described herein for providing modular, reusable shielding to storage containers. The system provides a flexible approach to allow expanding storage requirements while minimizing shielding needs. The RSP system is capable of shielding any number and configuration of containers while reducing the amount of shielding materials, reducing storage footprint, and allowing for simple reconfiguration. In some embodiments, the RSP system may be applied to the nuclear waste storage containers, including, for instance, In-Container Vitrification™ (ICV™) containers. FIG. 1 depicts a cross-section of an embodiment of an ICV™ container 399. Vitrification is the process by which a vitrified product with embedded contaminants is formed. Vitrification is the gold standard for long-term waste disposal due to the very low leachability of contamination out of the vitrified product. ICV™ is a system wherein the vitrification occurs in a one-time use or a reusable container. In some embodiments, the container is used only once for vitrification and serves as the final storage container. In some embodiments, a container may serve as the treatment and storage container for a vitrified waste form resulting from the treatment of solid wastes (ion-specific media (ISM), sludge, liquid processing waste, soils, ash, decontamination, and decommissioning wastes, etc.). The ICV™ container 399 depicted in FIG. 1 comprises outer shielding 457, refractory lining 431, feed port 411, starter path (not shown), electrodes 421, and lid (built in hood) 458. In some embodiments, the outer shielding 457 is composed of a metal such as steel. The lid 458 may comprise one or more electrode penetration/seal 415 assemblies that keep electrodes 421 in contact with the starter path while providing electrical insulation between the electrodes 421 and the ICV™ container 399. The ICV™ container 399 is described in more detail in Ion Specific Media Removal from Vessel for Vitrification, Ser. No. 15/012,101 filed Feb. 1, 2016, with a priority date of Feb. 1, 2015, which is hereby incorporated by reference in its entirety. The depicted embodiments show ICV™ containers as example storage containers. It should be clear that the containers are not necessarily ICV™ containers and may take other forms. The same principles and design aspects may be applied to many different styles and configurations of containers. The term “container” as used herein may refer to an ICV™ container or any other container type or style that may utilize the shielding principles and/or designs disclosed herein. While vitrified nuclear waste is disclosed as an example material requiring shielding in storage it should be clear that the same principles may be applied to other waste forms and other materials requiring shielding. For instance, in a temperature controlled facility the shielding may be used as thermal insulation. Electromagnetic shielding may be used for redirecting magnetic flux, and radio frequency shielding may be used to block radio waves. Other embodiments are contemplated. FIG. 2 depicts an isometric view of an embodiment of an ICV™ container 400. The depicted embodiment is a variation of the ICV™ container 399 depicted in FIG. 1, modified for the installation of a removable shield panel embodiment. The modifications comprise the addition of one or more shield mounting points 125 to facilitate mounting of shield panels. The shield mounting points 125 may vary in quantity, location, and form between various embodiments. Some embodiments of the shield panels may not require shield mounting points on the ICV™ container 400. In some embodiments, shield panels may be attached to the storage containers using one or more coupling mechanisms including magnetics, tongue and groove, suction cups, and Velcro®, among others. FIG. 3 depicts the modified ICV™ container 400 embodiment of FIG. 2 with shield mounts 125 and shield panels 100. Each container 400 may comprise one or more shield mounting points 125 on each side. In the depicted embodiment, each container 400 comprises two shield mounting points 125 on each side of the top of the container 400 for a total of eight shield mounting points 125 per container 400. The type, geometry, quantity, and location of the shield mounting points 125 may vary between embodiments. Shield mounts 150 are shaped to engage with the shield mounting points 125 on the container 400. In the depicted embodiment, a single container 400 is shielded on all sides. When containers 400 are stored they are generally stacked and layered. The internal containers 400 in a storage configuration often do not require individual shielding because shielding is at least partially provided by adjacent containers 400. When the containers 400 are stored together generally only the sides of the outermost containers 400 that are exposed to the storage environment require shielding. The RSP system may be used to shield external sides of stored containers thus reducing the amount of shielding required in a storage facility. As the number of containers 400 in a storage facility increases or decreases, the shielding of the outermost containers 400 may be easily adjusted by moving the removable shield panels 100 and preinstalling them on the exposed container 400 surfaces. FIGS. 4 and 5 depict single layer container 400 configurations where the shield panels 100 are mounted on only the outermost (exposed) surfaces of the containers 400 and secured with shield mounts 150. Top shield panels may be used to cover the top of the uppermost layer of containers 400. FIG. 6 depicts an example embodiment of a layer of ICV™ containers 400 containing vitrified nuclear waste. Nuclear waste is often classified by activity level with the common levels being low, intermediate, and high activity waste. Low activity waste generally requires little or no shielding whereas high activity waste may require a large amount of shielding. In the depicted embodiment, the containers 400 are filled with different classes of nuclear waste. The innermost container 400 is high (H) level and the surrounding containers 400 are intermediate (I) level. This embodiment illustrates how a lower level waste (the intermediate waste) can be used as shielding for higher level waste thus reducing shielding requirements in the storage facility. Reducing the amount of shielding reduces the storage footprint of each container 400 thus increasing capacity and efficiency of a storage facility. Additionally, the RSP system decouples the shielding from the container 400 from a weight standpoint thereby potentially increasing the amount of material that can be stored in each container 400. In some embodiments, the containers 400 may be stacked in two or more layers to minimize storage footprint and maximize storage capacity. FIGS. 7 and 8 depict ICV™ containers 400 in example stacked configurations with mounted removable shield panels 100 and top shield panels 200, secured with shield mounts 150. While the depicted embodiments comprise two layers it should be clear that the containers may be stored in other configurations included one or more layers. FIG. 9A depicts an embodiment of a generic removable shield panel 100. FIG. 9B depicts an example shield panel 100 a comprising tabbed edges 915 which may overlap to prevent gaps between shield panels 100 a when they are used side by side. Removable shield panels 100 may be composed of a wide range of materials which may be dependent upon the shielding's purpose. Shield panels 100 may vary in thickness and/or comprise layers of different materials. FIG. 9C depicts a top down cross-sectional view of an example shield panel 100 c comprising three layers 72, 73, and 74 of differing materials. Different embodiments may comprise varying numbers and thicknesses of layers of one or more different materials. For example, in nuclear waste storage, shield panels 100 may comprise one or more layers of materials including one or more of concrete, steel, lead, and mullite refractory, among others, to reduce radiation dosage rates. In some embodiments, steel shield panels have a half-value layer of 16 mm for Cs-137/Ba-137m radiation. Other half-value layer configurations are possible. In temperature-controlled facilities shield panels may comprise thermal insulation material(s). In some embodiments, shield panels may be composed of, or comprise a layer of, a bumper or impact resistant material to protect storage contents from impact. Shield panels may comprise conductive or magnetic materials, such as copper in some embodiments, to shield storage contents from electromagnetic flux. In some embodiments, shield panels may comprise multiple layers of differing materials operable to provide shielding of one or more different types. For example, electronic equipment may utilize shield panels that comprise at least a thermal shield layer and an electromagnetic shield layer. In some embodiments, one or more shield panels or materials therein may be layered wherein they connect using an interlocking concept similar to LEGOs® such that layers may be added and removed without modification to the shielding mounts. In some embodiments, one or more shield panels or materials therein may be layered wherein they connect using one or more of magnetism, suction, Velcro®, or other removable connection types known in the art. In some embodiments, such as the embodiment depicted in FIG. 10A, shield panels 100 may comprise circuitry 99 including temperature control mechanisms for providing cooling or heating to storage containers. In such embodiments, shield panels 100 may comprise electric circuit connectors 98 such that the connectors 98 align for simple connection during setup/reconfiguration. In some embodiments, such as for temporary storage and/or transportation, each shield panel may comprise standalone temperature control mechanisms. In some embodiments, shield panels may be hollow or comprise channels in the side facing the containers to reduce weight and/or to allow controlled airflow around the storage containers. In some embodiments, shield panels may comprise one or more sensors. Sensors may serve to alert in the event of leakage, temperatures outside acceptable ranges, vibration, radiation, and other conditions that may be detrimental to the stored materials, the environment, and/or workers. In some embodiments, the shield panels may further comprise one or more mechanisms to facilitate placement, lifting, and removal. The mechanisms may take the form of hooks, handles, recesses, and magnetic connectors, among others. The one or more mechanisms may, when not in use, lay flush with, recessed from, or protruding from the surface of the shield panel, in some embodiments. FIG. 10B depicts the removable shield panel embodiment of FIG. 9A further comprising example hooks, handles, and magnetic connectors to facilitate reconfiguration. The depicted placement facilitation mechanisms are shown for example purposes only. The particular combination, types, amount, positioning, geometry, and sizes of the depicted mechanisms may vary between embodiments. In the embodiment depicted in FIG. 10B, the shield panel 100 c comprises three example shield placement facilitation mechanisms: recesses 64, magnetic connectors 65, hook 66, and handles 67. The recesses 64 may provide surfaces in the shield panel 100 c upon which an upward force may be applied for lifting and repositioning the shield panel 100 c. Magnetic connectors 65 may provide areas or sections of the shield panel 100 c which are magnetic such that a magnetic force may be applied to lift and transport the shield panel 100 c. Hook 66 may be hinged such that it may fold upwards when needed to lift the shield panel 100 c and down against or recessed into the shield panel 100 c when not in use. Handles 67 may fold outwards or slide upwards from the shield panel 100 c as needed. In some embodiments, corner shielding may be provided along the edges to cover any gaps that may exist between side shield panels 100 (FIG. 9A) and between top panels and side shield panels 100 (FIG. 9A). FIG. 11 depicts an embodiment for example corner shielding types. Corner 815 shows overlapping side shield panels 100 secured to container 400 with shield mounts 150. Corner 845 shows corner shielding that may be used for tabbed shield panels (FIG. 9B). Corner 825 is a simple L shaped overlapping corner piece. Corner 805 is a simple square cross-section panel. Corner 835 is a combination of corner 805 and 825. In some embodiments, corner shielding may be attached to the shield panels using one or more coupling mechanisms including magnetics, tongue and groove, dovetail joints, suction cups, and Velcro®, among others. FIG. 12A depicts an alternate embodiment of a side shield 100 d with mounting points 131 for additional shield mounts 150 (FIG. 12B) on both the top and bottom sides of the shield 100 d for added stability and easier reconfiguration. In some embodiments, bottom shield mounts and mounting points 131 may be the same or similar geometry as top shield mounts and mounting points 130. FIG. 12B depicts the shield panel 100 d in use. In some embodiments, bottom shield mounts 131 may mount orthogonally or at an angle from the side rather than from the bottom such that they may be removed without having to lift or move the container. The addition of bottom mounts 131 may require a pull and lift force in order to remove the shield panels 100 d. Adding an extra force for removal increases stability, thus reducing chances of slippage over time or slippage due to outside forces or impacts such as earthquakes. FIG. 13 depicts an embodiment that utilizes bottom shields 201 in similar geometry as the top shields 200. In some embodiments, top shields 200 and bottom shields 201 may be incorporated with the side shields 100 to completely shield one or more containers. In some embodiments, the shield mount may be designed to secure a combination of one or more side shields 100, top shields 200, and bottom shield 201 together forming an enclosure for housing one or more containers. In some embodiments, bottom shielding is not required as the floor of the storage facility may provide adequate shielding. In some embodiments, bottom shielding may be in the form of a continuous pad or section of flooring. FIG. 14 depicts an embodiment of a shield mount 150. The depicted shield mount 150 comprises slots 124 and 126 where slot 124 fits over a mount point on the modified ICV™ container and slot 126 fits over a mount point in the shield panel. The slotted mounting mechanism facilitates simple mounting of shield panels and allows the shield panels to be easily lifted upwards for removal. When the shield mount 150 is placed correctly and completely the top surface 121 is flush with the top of the container and the outer surface 127 is flush with the outer surface of the shield panel, in some embodiments. In some embodiments, one or both of surface 121 and surface 127 may be either recessed or protruding. The filleted corners 120 allow for the shield mount 150 to be easily removed by hand or hand tool, if necessary. Typically the shield panels may be removed and reconfigured remotely. In some embodiments, one or more of the shield mounts 150 may be integrated with the shields. In some embodiments, a crane and/or robotic manipulator arm may be used as an apparatus for shield reconfiguration wherein the apparatus may be locally or remotely controlled. Some embodiments may utilize a robotic remote control system for shield reconfiguration. An example of such a robotic control system may be found in co-pending U.S. patent application Ser. No. 15/591,978, entitled System and Method for a Robotic Manipulator Arm, filed May 10, 2017, with a priority date of May 16, 2016, which is hereby incorporated by reference in its entirety elsewhere in this document. FIGS. 15A through 15C depict a variation of the shield mount embodiment of FIG. 14. The shield mount 150 a has many of the same features as the shield mount 150 depicted in FIG. 14. Shield mount 150 a has a closed slot 126 a where shield mount 150 (FIG. 14) has an open slot 126 (FIG. 14). FIG. 15B and FIG. 15C depict the shield mount 150 a in use. Closed slot 126 a fits over guide 112 in the shield panel 100. In the depicted embodiment, the shield mount 150 a is slidably attached to the shield panel 100 where slot 126 a slides along guide 112. FIGS. 15D and 15E depict an embodiment of a shield panel 100 corresponding to the shield mount embodiment of FIGS. 15A through 15C. In some embodiments, the shield mount 150 a may be fixed to the shield panel 100. The guide 112 keeps the shield mount 150 a aligned and prevents the shield mount 150 a from being separated from the shield panel 100. In FIG. 15B the shield mount 150 a is fully engaged with the shield panel 100 and the container 400. In FIG. 15C the shield mount 150 a is extended from the shield panel 100 and the container 400. The shield panel system allows for simple adjustment of shield thickness as necessary. For instance, in nuclear waste storage embodiments, shield thickness may require adjustment to maintain dose at acceptable limits (such as 1 mSv/hr on contact). In some embodiments, containers may be stored such that the higher activity containers are stored innermost and lower activity containers are stored outermost to increase shielding of the higher activity containers. If additional shielding is required the panels can be stacked to increase the shield thickness. FIG. 16A depicts an embodiment of an adjustable shield mount 500 that can be adjusted for different shield thicknesses. The positions of the shield mounting peg 530 and container mounting peg 520 can be adjusted by sliding them along the length of the cut channel 515 to compensate for varying shield thicknesses. In some embodiments, the shield mounting peg 530 and the container mounting peg 520 may be a single component. In the depicted embodiment nuts are used to tighten and secure the mounting pegs in position; however, other fastening mechanisms may be used. FIG. 16B depicts the adjustable shield mount 500 in use with a thick shield 100 e. In some embodiments, the adjustable shield mount 500 may further comprise a toggle clamp or other such clamping or securing mechanism. FIG. 17A depicts an embodiment of an adjustable shield mount 550 that can accommodate two shields of different thicknesses. The positions of both shield mounting pegs 530 and the container mounting peg 520 can be adjusted by sliding them along the length of the cut channel 515 to compensate for varying shield thicknesses. In some embodiments, the container mounting peg and the nearest shield mounting peg 530 may be a single component. FIG. 17B depicts the adjustable shield mount 550 in use with two shield panels 100. In the depicted embodiment, the shield panels 100 are the same thickness; however, they may be different thicknesses in other embodiments. In the depicted embodiment, nuts are used to tighten and secure the mounting pegs in position; however, other fastening mechanisms may be used. In some embodiments, the adjustable shield mount 550 may further comprise a toggle clamp or other such clamping or securing mechanism. FIG. 18 depicts an example embodiment of a layer of ICV™ containers 400 containing vitrified nuclear waste. In the depicted embodiment, the containers 400 are filled with different classes of nuclear waste. Those marked H contain high level waste and those marked I contain intermediate level waste. Generally in storage configurations containing different waste levels the lower level waste may be used as shielding for the higher level waste, such as the example embodiment depicted in FIG. 6. When it is not possible to use the lower level waste as additional shielding against the higher level waste different types, thickness, and/or layers of shielding may be needed on the higher level waste than on the lower level waste. In the depicted embodiment, all of the same shields are used; however, the shielding is doubled on the higher-level waste. This is an example of when it is useful to have adjustable shield mounts capable of accommodating different numbers and thicknesses of shield panels 100. FIGS. 19A through 19C are described as a group. FIG. 19A depicts an isometric view of an embodiment of a shield mount 150 a that utilizes a toggle clamp mechanism 300 to secure a modified shield panel 100 f (FIG. 19C). The toggle clamp 300 fits over the base of the shield mount 150 a and allows the shield mount 150 a to be secured to the shield panel 100 f and the top of the ICV™ container 400 with a clamp mechanism 300 to prevent the shield from slipping downward. FIG. 19B depicts a side view of the shield mount 150 a. FIG. 19C depicts the shield mount 150 a in use with modified shield panels 100 f. In an embodiment, the size and materials used for clamp mechanism 300 may vary based on the size and composition of the shield panel. It should be clear that a 5000-pound shield panel may require sturdier and larger materials for clamp mechanism 300 than a 100-pound shield panel. FIG. 20 depicts an embodiment of an adjustable shield mount 500 a that incorporates the use of a toggle clamp system 300 for securing shield panels of varying thicknesses. The depicted embodiment incorporates the shield mounting peg 530 and the container mounting peg 520 to accommodate shield panels of varying thicknesses. FIG. 21A depicts an embodiment of a shield mount 700 that secures the side shields 100 f (FIG. 21C) with a top shield 200 a (FIG. 21B). The shield mount 700 comprises an edge gripper 720 which may be used to secure the top shield 200 a (FIG. 21B) in place via clamping force and friction. In the depicted embodiment, the edge gripper 720 is fastened to the end of the shield mount 700. In some embodiments, the edge gripper 720 may be integrated to the shield mount 700. In some embodiments, the shield mount may be integrated to the top shield 200. FIG. 21B depicts an embodiment of a top shield 200 a that couples with the shield mount embodiment 700. FIG. 21C depicts the top shield embodiment 200 a in use with the shield mount 700 and shield panel 100 f. In some embodiments, the top shield 200 a is sized to fit just over the container lid. Figures, figure elements, and written disclosure related to the following embodiment are described in detail in the above disclosure. The RSP system allows for modular reconfigurable shielding for one or more storage containers. In an example embodiment, there are a plurality of unshielded storage containers containing nuclear waste. In industry, any container for storing nuclear waste normally comprises, as part of its structure (i.e. not removable), the required shielding for the particular waste level contained therein to keep the radiation dosage below predetermined safety limits. In this example embodiment the nuclear waste storage containers are unshielded i.e. they can be used to store any level of nuclear waste because the shielding required for a particular waste level is not included as part of their structure. These unshielded nuclear waste storage containers are modular and reconfigurable because they can contain any waste level and appropriate shielding can be added as needed based on predetermined dosage requirements for a given storage facility. In the example embodiment, each unshielded nuclear waste storage container comprises at least one mounting point for mounting one or more modular shield panels to it. Each modular shield panel comprises at least one mounting point for mounting to an unshielded nuclear waste storage container. Depending on the number of shield panels required and the number of shield mounts on the shield panels and the containers, one or more shield mounts may be used to couple with the mounting points on the shield panels and the containers to attach the shield panels to the containers. In some embodiments, one or more of the shield mounts may be adjustable to accommodate shield panels of varying thicknesses. In the example embodiment, a plurality of nuclear waste storage containers may be stored together. When stored together the sides adjacent to (face-to-face with) other storage containers do not require shielding while, depending on the waste levels contained therein, and the predetermined dosage requirements for the particular storage facility, the outermost (external) faces of the storage containers may require shielding. The sides of the containers that are placed adjacent to other containers do not require additional shielding because the shielding on that side is provided by the neighboring container. Continuing with the example embodiment, when the storage containers are placed in a storage configuration and all of the external facing sides of the containers are shielded according to the requirements of the particular waste level and/or storage facility the storage configuration is considered to be fully shielded. When an additional unshielded storage container needs to be added to the storage configuration, depending on the layout of the existing configuration, one or more shield panels may be removed from one or more storage containers in the configuration resulting in one or more partially shielded storage containers. The additional unshielded storage container may then be placed in the configuration adjacent to one or more partially shielded storage containers in the configuration. One or more of the previously removed one or more shield panels may then be installed on the external faces of the newly added storage container. If any faces are still exposed (unshielded) additional shield panels may be installed as needed to result in a fully shielded storage configuration. It should be clear that any one or more aspects of the disclosed shield panels, shield mounts, and shielding configurations may be combined to form other embodiments not expressly disclosed herein. Additionally, the shield mounts may take other geometries and utilize fasteners different than those depicted. For the sake of convenience, the operations are described as various interconnected functional blocks or distinct software modules. However, this is not necessary, and there may be cases where these functional blocks or modules are equivalently aggregated into a single logic device, program or operation with unclear boundaries. In any event, the functional blocks and software modules or described features can be implemented by themselves, or in combination with other operations in either hardware or software. Having described and illustrated the principles of the systems, methods, processes, and/or apparatuses disclosed herein in a preferred embodiment thereof, it should be apparent that the systems, methods, processes, and/or apparatuses may be modified in arrangement and detail without departing from such principles. Claim is made to all modifications and variation coming within the spirit and scope of the following claims.
	claims	1. An apparatus for treating a radioactive nitrate waste liquid, comprising:a denitrification tank that accommodates active sludge which adsorbs or takes in a radioactive substance in a nitrate waste liquid containing nitrate and the radioactive substance and in which an anaerobic microorganism that reduces the nitrate to nitrogen gas grows; anda reaeration tank that aerates and mixes a denitrification-treated liquid treated in the denitrification tank with active sludge in which the aerobic microorganism grows, whereinthe denitrification tank includes:a pH adjusting unit that supplies a pH adjuster used for adjusting pH of the nitrate waste liquid;a carbon-source supplying unit that supplies a carbon source to the denitrification tank;a first solid-liquid separating unit that separates a denitrified liquid treated with the active sludge into a sludge-containing solid content and a denitrification-treated liquid; anda gas supplying unit that is arranged on a lower side of the first solid-liquid separating unit and supplies gas not containing oxygen into the denitrification tank, andthe reaeration tank includes:a second solid-liquid separating unit that further separates the denitrification-treated liquid treated with the active sludge into a sludge-containing solid content and a treated liquid; andan air supplying unit that is arranged on a lower side of the second solid-liquid separating unit and supplies air into the reaeration tank. 2. The apparatus for treating a radioactive nitrate waste liquid according to claim 1, wherein pH of the denitrified liquid in the denitrification tank is from 7.0 to 10.0. 3. The apparatus for treating a radioactive nitrate waste liquid according to claim 1, wherein the denitrification tank includes a gas circulating line through which nitrogen gas and carbon dioxide gas produced by a reaction between the anaerobic microorganism and the nitrate in the denitrified liquid are circulated into the denitrification tank. 4. The apparatus for treating a radioactive nitrate waste liquid according to claim 1, further comprising:a pH sensor that measures pH of the denitrified liquid; anda carbon-dioxide-gas supply-amount adjusting valve that adjusts an amount of carbon dioxide gas supplied to the denitrification tank. 5. The apparatus for treating a radioactive nitrate waste liquid according to claim 1, whereina carbon source supplied to the denitrification tank is acetic acid, andthe apparatus further includes:a pH sensor that measures pH of the denitrified liquid; anda carbon-source supply-amount adjusting valve that adjusts an amount of a carbon source supplied to the denitrification tank. 6. The apparatus for treating a radioactive nitrate waste liquid according to claim 3, whereina carbon source supplied to the denitrification tank is one of or both of organic acid and sweetener, andair is temporarily supplied to the gas circulating line via an air-supply-amount adjusting valve. 7. The apparatus for treating a radioactive nitrate waste liquid according to claim 6, wherein the denitrification tank includes an oxidation-reduction potential meter that measures an oxidation-reduction potential of the denitrified liquid. 8. The apparatus for treating a radioactive nitrate waste liquid according to claim 1, comprising an adjusting tank at an upstream of the denitrification tank, whereinthe adjusting tank includes:an electrical conductivity meter that measures a level of electric conductivity;an industrial-water introducing line; andan industrial-water supply-amount adjusting valve that is interposed through the industrial-water introducing line and adjusts an amount of water supplied based on a measurement value of the electrical conductivity meter. 9. The apparatus for treating a radioactive nitrate waste liquid according to claim 1, wherein the first solid-liquid separating unit and the second solid-liquid separating unit are solid-liquid separating films. 10. An apparatus for treating a radioactive nitrate waste liquid, comprising:a denitrification tank that accommodates active sludge which adsorbs or takes in a radioactive substance in a nitrate waste liquid containing nitrate and the radioactive substance and in which an anaerobic microorganism that reduces the nitrate to nitrogen gas grows; anda reaeration tank that aerates and mixes a denitrification-treated liquid treated in the denitrification tank with active sludge in which the aerobic microorganism grows, whereinthe denitrification tank includes:a pH adjusting unit that supplies a pH adjuster used for adjusting pH of the nitrate waste liquid;a carbon-source supplying unit that supplies a carbon source to the denitrification tank;a weir that prevents flowing-out sludge from leaking out into the denitrification tank, andthe reaeration tank includes:a second solid-liquid separating unit that further separates the denitrification-treated liquid treated with the active sludge into a sludge-containing solid content and a treated liquid; andan air supplying unit that is arranged on a lower side of the second solid-liquid separating unit and supplies air into the reaeration tank. 11. A method for treating a radioactive nitrate waste liquid, comprising:in a denitrification tank that accommodates active sludge that adsorbs or takes in a radioactive substance in a nitrate waste liquid containing nitrate and the radioactive substances and in which an anaerobic microorganism that reduces the nitrate to nitrogen gas grows, a step at which an active sludge treatment is performed by supplying a carbon source while adjusting pH of the nitrate waste liquid, and at a time of separating a denitrified liquid treated with the active sludge by using a first solid-liquid separating unit into a sludge-containing solid content and a denitrification-treated liquid, gas not containing oxygen is supplied thereby to accelerate a process for reducing an anaerobic microorganism and cleanse the active sludge adhered to the solid-liquid separating unit with the gas; andin a reaeration tank that aerates and mixes the denitrification-treated liquid treated in the denitrification tank with the active sludge, a step of further separating the denitrification-treated liquid treated with the active sludge into a sludge-containing solid content and a reaeration-treated liquid by using a second solid-liquid separating unit.
042736145	summary	BACKGROUND OF THE INVENTION The present invention relates to a handling device for an assembly, particularly for a fertile or fissile assembly of a fast neutron nuclear reactor with cooling by a liquid metal. It is known that in such a reactor the assemblies constituting the reactor core are formed by open enclosures containing a bundle of sheathed needles of fertile or fissile material. These assemblies are juxtaposed and rest in a vertical position on a supporting member. The complete core assembly is submerged in a suitable volume of liquid metal, generally sodium, contained in a container which is open at its upper end and is suspended beneath a thick slab, particularly made from concrete which extends horizontally and encloses a protective caisson surrounding the container. The liquid sodium circulates from bottom to top in the enclosures of the assemblies in contact with the needles in such a way as to extract the calories given off by the latter, the heated sodium being taken up by heat exchangers and then on leaving the latter by pumps which return the cooled sodium beneath the supporting member for a further passage through the core. In operation, the mass of liquid sodium in the container generally reaches a given level, separated from the lower face of the slab by an inert gas, generally argon cushion. At given intervals, variable according to the reactor operating conditions, it is necessary to replace the assemblies in the core. For this purpose, the installation has transfer means controlled from the outside of the protective caisson and which after traversing the slab are able to take up and extract one by one the assemblies in the core, whereby said assemblies are laterally displaced whilst maintaining them below the sodium level and finally they are deposited in a handling pot located at a loading and unloading station located on the side of the core and which is also beneath the sodium level. The handling pot is generally formed by an elongated receptacle open at its upper end via which the assembly is introduced and closed at its inner base, said pot generally being provided with means which, during its movement to an unloading station, permit the distribution and evacuation in acceptable proportions of the residual calories given off by the assembly. In a generally adopted construction, the handling pot is placed on a carriage against which it bears whereby after tilting said carriage is able to move in an inclined ramp by means of a cable or pull chain, thereby making it possible to remove the carriage, handling pot and the assembly contained therein from the container and caisson in order to bring them in tightly sealed manner in a chamber provided externally of said caisson to a position above the slab where the carriage is tilted and then returned over another inclined ramp. It then once again traverses the slab, externally of the caisson, and issues into an appropriate storage container where the pot is again submerged beneath an appropriate volume of liquid sodium and finally arrives at a loading and unloading station. Advantageously, the attachment straps for the carriage are provided with a parachute brake system, which prevents the sudden dropping of the pot containing the assembly in the case of an accidental breakage of the pull cable. Such a device is in particular described and claimed in French patent application No. 2,192,057. When the carriage and the pot are in a position vertical to the loading and unloading station of the reactor or storage container the attachment straps, whose lower end is connected to the carriage and whose upper end is connected to the parachute brake which always remains on the inclined ramp, are disposed obliquely and provide vertical access to the pot in order to permit the introduction or removal of an assembly. During an unloading operation the handling pot is initially completely submerged below the sodium level. When the carriage arrives in the vicinity of the slab, the upper end of the pot emerges above this level into the inert gas cushion and then the whole pot progressively emerges so that it is totally penetrating the gaseous atmosphere in the container above the sodium level and communicates with the inside of the chamber. Under these conditions the pot containing the assembly must be closed so as to prevent the evaporation of sodium, which may occur in the case of an accidental immobilisation of the pot in a partly or completely emerged position as a result of the rise in the sodium temperature. The drop in the sodium level may cause the circulation by convection of the sodium in contact with the assembly to stop, resulting in a local temperature rise which may even bring about a fusion of the fuel, which constitutes an inadmissible accident at the safety level. BRIEF SUMMARY OF THE INVENTION The present invention relates to a device for handling an assembly comprising a handling pot and a closing device ensuring the sealing of said pot in an automatic manner and without any mechanical action as soon as the upper end of the pot emerges from the sodium. To this end, the present device intended for a fast neutron reactor cooled by a liquid metal such as sodium and of the type comprising a container with a vertical axis containing the core and a given volume of sodium contained in said container, defining above the core a free level surmounted by an inert gas cushion, a thick horizontal slab placed above the gas cushion and constituting the upper closure of a protective caisson surrounding the container, an inclined ramp which passes into the container up to a loading and unloading station located in the vicinity of the core and extending at its opposite end into a handling chamber supported by the slab, a second inclined ramp connecting the inside of the handling chamber to a second loading and unloading station positioned externally of the caisson in a storage container, a carriage carrying a handling pot able to move along the said ramps and moving the pot from a position where it is submerged below the sodium level to a position where it is introduced into the inert gas atmosphere above the sodium level, said pot being open at its upper end so as to permit the introduction or removal of the assembly to be handled when it is positioned vertically on its carriage with respect to said loading and unloading stations provided for this purpose in the reactor vessel and in the storage container, wherein during its displacement between said stations the carriage is associated, to the right of the upper open end of the pot, with a generally cylindrical enclosure which is open towards the bottom and in which is mounted a hollow member, whereof at least the lower part has a spherical shape, whereby the hollow member bears via the spherical lower part against the upper open end of the pot when the latter emerges above said level. According to a first embodiment, the said member is constituted by a sphere. According to a second embodiment, said member is constituted by a hemisphere having a planar upper face. In a constructional variant of the device, the hollow member bears against the upper part of the enclosure when the pot and its carriage are submerged on the inclined ramps. According to another variant, the hollow member also bears against the upper end of the pot when the latter and its carriage are submerged on the inclined ramps. According to a special feature of the device, the upper open end of the handling pot has a conical bearing surface against which is applied the spherical part of the hollow member when said end emerges above the sodium level. Advantageously, the hollow member is filled with an inert gas. The interior of the hollow member may also be maintained under vacuum. According to a special embodiment of the invention, facing the open end of the pot the cylindrical enclosure has a contraction towards the inside of the enclosure able to retain the hollow member when there is no pot on the carriage. According to a special embodiment of the invention, the upper part of the enclosure is provided with orifices. According to another variant, the cylindrical enclosure laterally has longitudinal elongated openings diametrically opposed in pairs and in which are engaged transverse pins provided in the lateral surface of the spherical member.
	abstract	The invention relates to a method for determining lens errors in a Scanning Electron Microscope, more specifically to a sample that enables such lens errors to be determined. The invention describes, for example, the use of cubic MgO crystals which are relatively easy to produce as so-called ‘self-assembling’ crystals on a silicon wafer. Such crystals have almost ideal angles and edges. Even in the presence of lens errors this may give a clear impression of the situation if no lens errors are present. This enables a good reconstruction to be made of the cross-section of the beam in different under- and over-focus planes. The lens errors can then be determined on the basis of this reconstruction, whereupon they can be corrected by means of a corrector.
	summary
042467837	summary	The invention relates to a measuring device for determining and checking the spring force of the resilient spacer projections that are set into spacer grids for nuclear reactor fuel assemblies and which push the fuel rods surrounded by each spacer mesh against, usually, two oppositely disposed rigid spacer projections. Nuclear reactor fuel assemblies generally include a multiplicity of fuel rods disposed parallel to each other which are held at the desired nominal distance from each other by spacer grids. These spacer grids, in turn, are normally constructed from crossed sheet metal webs. The meshes formed thereby surround one fuel rod each. For centering this fuel rod, resilient and rigid contact projections are provided in the respective mesh, where normally, as seen in the axial direction of the fuel rod, a resilient contact projection in one spacer wall is disposed opposite two rigid contact projections in the spacer wall parallel thereto. This method of centering not only permits easy insertion of the fuel rods into the spacer meshes but, during the operation of the reactor, also ensures maximum freedom from vibration of the fuel rods, around which the reactor coolant flows with high velocity. The resilient contact projections are of particular importance here. Their spring force, in addition to the accuracy of the physical dimensions, must be constant within narrow tolerances, so that the above-mentioned functions of the spacer grid remain assured, particularly with respect to the vibration behavior. The problem therefore arose to overcome the disadvantages associated with the previously used inaccurate and difficult to use measuring methods and to find a simple-to-operate measuring device to be used, on the one hand, during the manufacture of the spacer grids and, on the other hand, for checking spacer grids already subjected to reactor operation, which allows measurement and checking of the spring force of the resilient contact projections. It is therefore an object of the invention to provide a spring-force measuring device which overcomes the hereinaforementioned disadvantages of the heretofore known devices of this general type. With the foregoing and other objects in view, there is provided, in accordance with the invention, a device for measuring the spring-force of resilient spacer projections set into spacer grids of nuclear reactor fuel assemblies to push fuel rods surrounded by spacer meshes against at least two oppositely disposed rigid spacer projections, comprising a force measuring plug having a diameter equal to the diameter of a fuel rod to be fixed in the spacer grid, and a flexible beam integral with the force measuring plug, the flexible beam, having a free end in contact with a first resilient spacer projection to be measured, and another end firmly connected to the force measuring plug and having at least one wire strain gage disposed thereon. In accordance with another feature of the invention, the flexible beam is formed by a stress-free slot formed in the force measuring plug. In accordance with a further feature of the invention, a second resilient spacer projection is disposed at a 10.degree. shift from the first resilient spacer projection in the spacer grid, said slot forming the flexible beam being formed step-like whereby the second resilient spacer projection contacts the force measuring plug and not the flexible beam. In accordance with a concomitant feature of the invention, the force measuring plug is hollow and has a lateral hole formed therein, the other end of the flexible beam being centered within the hollow force measuring plug, and including a pin disposed on the free end of the flexible arm extending radially through the lateral hole, the pin having a rounded end contactable with the first resilient spacer projection. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a spring-force measuring device, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims.
	description	The field of invention relates generally to the field of semiconductor integrated circuit manufacturing and, more specifically but not exclusively, relates to correction of systematic non-uniformities using a gas cluster ion beam. Gas-cluster ion beams (GCIB's) are used for etching, cleaning, smoothing, and forming thin films. For purposes of this discussion, gas clusters are nano-sized aggregates of materials that are gaseous under conditions of standard temperature and pressure. Such gas clusters may consist of aggregates including a few to several thousand molecules, or more, that are loosely bound together. The gas clusters can be ionized by electron bombardment, which permits the gas clusters to be formed into directed beams of controllable energy. Such cluster ions each typically carry positive charges given by the product of the magnitude of the electronic charge and an integer greater than or equal to one that represents the charge state of the cluster ion. The larger sized cluster ions are often the most useful because of their ability to carry substantial energy per cluster ion, while yet having only modest energy per individual molecule. The ion clusters disintegrate on impact with the workpiece. Each individual molecule in a particular disintegrated ion cluster carries only a small fraction of the total cluster energy. Consequently, the impact effects of large ion clusters are substantial, but are limited to a very shallow surface region. This makes gas cluster ions effective for a variety of surface modification processes, but without the tendency to produce deeper subsurface damage that is characteristic of conventional ion beam processing. Conventional cluster ion sources produce cluster ions having a wide size distribution scaling with the number of molecules in each cluster that may reach several thousand molecules). Clusters of atoms can be formed by the condensation of individual gas atoms (or molecules) during the adiabatic expansion of high-pressure gas from a nozzle into a vacuum. A skimmer with a small aperture strips divergent streams from the core of this expanding gas flow to produce a collimated beam of clusters. Neutral clusters of various sizes are produced and held together by weak interatomic forces known as Van der Waals forces. This method has been used to produce beams of clusters from a variety of gases, such as helium, neon, argon, krypton, xenon, nitrogen, oxygen, carbon dioxide, sulfur hexafluoride, nitric oxide, nitrous oxide, and mixtures of these gases. Several emerging applications for GCIB processing of workpieces on an industrial scale are in the semiconductor field. Although GCIB processing of workpieces is performed using a wide variety of gas-cluster source gases, many of which are inert gases, many semiconductor processing applications use reactive source gases, sometimes in combination or mixture with inert or noble gases, to form the GCIB. Although gas cluster ion beams have been used to correct for variations in an upper layer of a workpiece, gas cluster ion beams have not been used to correct non-uniformities caused by tool specific process anomalies that repeatedly affect the output process parameters of a GCIB apparatus. An apparatus and method for correcting non-uniformities, including non-uniformities of an upper layer of an incoming workpiece and systematic non-uniformities caused by inherent variation in a process tool such as a gas cluster ion beam tool, using a gas cluster ion beam is disclosed in various embodiments. However, one skilled in the relevant art will recognize that the various embodiments may be practiced without one or more of the specific details, or with other replacement and/or additional methods, materials, or components. In other instances, well-known structures, materials, or operations are not shown or described in detail to avoid obscuring aspects of various embodiments of the invention. Similarly, for purposes of explanation, specific numbers, materials, and configurations are set forth in order to provide a thorough understanding of the invention. Nevertheless, the invention may be practiced without specific details. Furthermore, it is understood that the various embodiments shown in the figures are illustrative representations and are not necessarily drawn to scale. In the description and claims, the terms “coupled” and “connected,” along with their derivatives, are used. It should be understood that these terms are not intended as synonyms for each other. Rather, in particular embodiments, “connected” may be used to indicate that two or more elements are in direct physical or electrical contact with each other while “coupled” may further mean that two or more elements are not in direct contact with each other, but yet still co-operate or interact with each other. Reference throughout this specification to “one embodiment” or “an embodiment” means that a particular feature, structure, material, or characteristic described in connection with the embodiment is included in at least one embodiment of the invention, but do not denote that they are present in every embodiment. Thus, the appearances of the phrases “in one embodiment” or “in an embodiment” in various places throughout this specification are not necessarily referring to the same embodiment of the invention. Furthermore, the particular features, structures, materials, or characteristics may be combined in any suitable manner in one or more embodiments. Various additional layers and/or structures may be included and/or described features may be omitted in other embodiments. Various operations will be described as multiple discrete operations in turn, in a manner that is most helpful in understanding the invention. However, the order of description should not be construed as to imply that these operations are necessarily order dependent. In particular, these operations need not be performed in the order of presentation. Operations described may be performed in a different order than the described embodiment. Various additional operations may be performed and/or described operations may be omitted in additional embodiments. There is a general need for correcting non-uniformities using a gas cluster ion beam. By correcting non-uniformities, including non-uniformities of an upper layer of an incoming workpiece and systematic non-uniformities caused by inherent variation in a process tool such as a gas cluster ion beam tool, a more repeatable and constant process may be provided. With reference to FIG. 1, a GCIB processing apparatus 200 includes a vacuum vessel 102 divided into communicating chambers that include a source chamber 104, an ionization/acceleration chamber 106, and a processing chamber 108 separated from the source chamber 104 by the ionization/acceleration chamber 106. The chambers 104, 106, 108 are evacuated to suitable operating pressures by vacuum pumping systems 146a, 146b, and 146c, respectively. A condensable source gas 112 (for example, argon (Ar), carbon dioxide (CO2), oxygen (O2), or nitrogen (N2)) stored in a source gas cylinder 111 is admitted under pressure through a gas metering valve 113 and a gas feed tube 114 into a stagnation chamber 116. The source gas is subsequently ejected from the stagnation chamber 116 into the substantially lower pressure vacuum inside the source chamber 104 through a properly shaped nozzle 110. A gas jet 118 results inside the source chamber 104. Cooling, which results from the rapid expansion of the jet 118, causes a portion of the gas jet 118 to condense into clusters, each consisting of from several to several thousand weakly bound atoms or molecules. A gas skimmer aperture 120 situated between the source chamber 104 and ionization/acceleration chamber 106 partially separates any gas molecules that have not condensed into clusters from those that have condensed and become part of the gas jet 118. The removal of the un-condensed gas molecules minimizes pressure increases in the downstream regions where such higher pressures would be detrimental, such as in the ionization/acceleration chamber 106 near ionizer 122 and high voltage electrodes 126 and in the process chamber 108. After the gas jet 118 has been formed in the source chamber 104, the constituent gas clusters in gas jet 118 are ionized by ionizer 122. The ionizer 122 is typically an electron impact ionizer that produces electrons from one or more filaments 124 and accelerates and directs the electrons causing them to collide with the gas clusters in the gas jet 118 inside the ionization/acceleration chamber 106. The electron impact ejects electrons from molecules in the gas clusters to generate ionized molecules and thereby endows the gas clusters with a net positive charge to define cluster ions. A filament power supply 136 provides voltage VF to heat the ionizer filament 124. A set of suitably biased high voltage electrodes 126 in the ionization/acceleration chamber 106 extracts the cluster ions from the ionizer 122. The high voltage electrodes 126 then accelerate the extracted cluster ions to a desired energy and focus them to define the GCIB 128. The kinetic energy of the cluster ions in GCIB 128 typically ranges from 1 thousand electron volts (keV) to several tens of keV. Anode power supply 134 provides voltage VA to at least one of the high voltage electrodes 126 for accelerating electrons emitted from filament 124 and causing the electrons to bombard the gas clusters in gas jet 118, which produces cluster ions. Extraction power supply 138 provides voltage VE to bias at least one of the high voltage electrodes 126 to extract ions from the ionizing region of ionizer 122 and to form the GCIB 128. Accelerator power supply 140 provides voltage VAcc to bias one of the high voltage electrodes 126 with respect to the ionizer 122 so as to result in a total GCIB acceleration energy equal to VACC electron volts (eV). Lens power supplies 142, 144 may be provided to bias some of the high voltage electrodes 126 with potentials (e.g., VL1 and VL2) to focus the GCIB 128. A beam filter 256 in the ionization/acceleration chamber 106 eliminates monomers or monomers and light cluster ions from the GCIB 128 to define a GCIB 202 that enters the processing chamber 108. A beam gate 222 is disposed in the path of GCIB 128 in the ionization/acceleration chamber 106. Beam gate 222 has an open state in which the GCIB 128 is permitted to pass from the ionization/acceleration chamber 106 to the processing chamber 108 to define GCIB 202 and a closed state in which the GCIB 128 is blocked from entering the processing chamber 108. A control cable 224 conducts control signals from dosimetry processor 214 to beam gate 222. The control signals controllably switch beam gate 222 to between the open or closed states. A workpiece 210, which may be a semiconductor wafer or other substrate to be processed by GCIB processing, is disposed in the path of the GCIB 202 in the processing chamber 108. Because most applications contemplate the processing of large workpieces 210 with spatially uniform results, a scanning system may be desirable to uniformly scan the GCIB 202 across large areas to produce spatially homogeneous results. The GCIB 202 directed at the workpiece 210 may be substantially stationary (i.e., un-scanned). Workpiece 210 is held in the processing chamber 108 on a X-Y positioning table 204 operable to move the workpiece 210 in two axes, effectively scanning the workpiece 210 relative to the GCIB 202. The GCIB 202 impacts the workpiece 210 at a projected impact region 244 on a surface of the workpiece 210. By X-Y motion, the X-Y positioning table 204 can position each portion of a surface of the workpiece 210 in the path of GCIB 202 so that every region of the surface may be made to coincide with the projected impact region 244 for processing by the GCIB 202. An X-Y controller 216 provides electrical signals to the X-Y positioning table 204 through an electrical cable 218 for controlling the position and velocity in each of X-axis and Y-axis directions. The X-Y controller 216 receives control signals from, and is operable by, system controller 228 through an electrical cable 226. X-Y positioning table 204 moves by continuous motion or by stepwise motion according to conventional X-Y table positioning technology to position different regions of the workpiece 210 within the projected impact region 244. In one embodiment, X-Y positioning table 204 is programmably operable by the system controller 228 to scan, with programmable velocity, any portion of the workpiece 210 through the projected impact region 244 for GCIB processing by the GCIB 202. Alternatively, orthogonally oriented electrostatic scan plates 130, 132 can be utilized to produce a raster or other scanning pattern of the GCIB 202 across the desired processing area on workpiece 210, instead of or in addition to using positioning table 204. When beam scanning is performed, a scan generator 131 provides X-axis and Y-axis scanning signal voltages to the scan plates 130, 132. The scanning signal voltages provided to the scan plates 130, 132 may be triangular waves of different frequencies that cause the GCIB 202 to scan the entire surface of workpiece 210. The workpiece holding surface 260 of positioning table 204 is electrically conductive and is connected to a dosimetry processor 214 by an electrical lead 212. An electrically insulating layer 258 of positioning table 204 isolates the workpiece 210 and workpiece holding surface 260 from the other portions of the positioning table 204. Electrical charge induced in the workpiece 210 by the impinging GCIB 202 is conducted through workpiece 210, workpiece holding surface 260, and electrical lead 212 to the dosimetry processor 214 for measurement. Dosimetry processor 214 has integrating means for integrating the GCIB current to determine a GCIB processing dose. Under certain circumstances, a target-neutralizing source (not shown) of electrons, sometimes referred to as electron flood, may be used to neutralize the GCIB 202. In such case, a Faraday cup (not shown) may be used to assure accurate dosimetry despite the added source of electrical charge. The processing chamber 108 includes optical windows 230 and 232. An optical transmitting transducer 234, which may also have additional transmitting optics 236, and an optical receiving transducer 238, which may also have additional receiving optics 240, form a conventional optical instrumentation system. The transmitting transducer 234 receives, and is responsive to, controlling electrical signals from the system controller 228 communicated through an electrical cable 246. The transmitting transducer 234 directs an optical beam through the optical window 230 toward the workpiece 210. The receiving transducer 238 detects the optical beam through optical window 232. The receiving transducer 238 sends measurement signals to the system controller 228 through an electrical cable 242. The optical instrumentation system may comprise any of a variety of different instruments for tracking the progress of the GCIB processing. For example, the optical instrumentation system may constitute a spectroscopic ellipsometry system for measuring or mapping the thickness of the upper film layer on the workpiece 210. As another example, the optical instrumentation system may comprise a scatterometer for measuring or mapping the thickness of the layer on the workpiece surface. By operating under control of the system controller 228 and in conjunction with the X-Y positioning table 204, the optical instrumentation can map one or more characteristics of the workpiece 210. In addition to gas cylinder 112, the GCIB processing apparatus 200 has a second gas cylinder 250 for containing a reactive gas 252, which may be, for example, oxygen, nitrogen, carbon dioxide, nitric oxide, nitrous oxide, another oxygen-containing condensable gas, or sulfur hexafluoride. Shut-off valves 246 and 248 are operable by signals transmitted through electrical cable 254 by system controller 228 to select either source gas 112 or source gas 252 for GCIB processing. The dosimetry processor 214 may be one of many conventional dose control circuits that are known in the art and may include, as a part of its control system, all or part of a programmable computer system. The X-Y controller 216 may include as part of its logic all, or part of, a programmable computer system. The dosimetry processor 214 may include as part of its logic all, or part of, a programmable computer system. Some or all of the logic of the X-Y controller 216 and dosimetry processor 214 may be performed by a small general purpose computer that also controls other portions of the GCIB processing apparatus 200, including the system controller 228. In operation, the dosimetry processor 214 signals the opening of the beam gate 222 to irradiate the workpiece 210 with the GCIB 202. The dosimetry processor 214 measures the GCIB current collected by the workpiece 210 to compute the accumulated dose received by the workpiece 210. When the dose received by the workpiece 210 reaches a predetermined required dose, the dosimetry processor 214 closes the beam gate 222 and processing of the workpiece 210 is complete. The dosimetry processor 214 is electrically coupled with the system controller 228 by an electrical cable 220. During processing of the workpiece 210, the dose rate is communicated by the dosimetry processor 214 to the system controller 228 by electrical signals transmitted over electrical cable 220. The system controller 228 analyzes the electrical signals to, for example, confirming that the GCIB beam flux is substantially constant or to detect variations in the GCIB beam flux. The X-Y controller 216 is responsive to electrical signals from the system controller 228 that are transmitted over an electrical cable 226. The X-Y controller 216 can scan the X-Y positioning table to position every part of the workpiece 210 for processing according to predetermined velocities that result appropriate beam dwell times to etch material or to deposit material to the desired local thicknesses effective to provide a film of substantially uniform thickness. As an alternative method, the GCIB beam 202 may be scanned at a constant velocity in a fixed pattern across the surface of the workpiece 210, but the GCIB intensity is modulated (often referred to as Z-axis modulation) to deliver an intentionally non-uniform dose to the sample. The GCIB intensity may be modulated in the GCIB processing apparatus 200 by any of a variety of methods, including varying the gas flow from a GCIB source supply; modulating the ionizer by either varying a filament voltage VF or varying an anode voltage VA; modulating the lens focus by varying lens voltages VL1 and/or VL2; or mechanically blocking a portion of the gas cluster ion beam with a variable beam block, adjustable shutter, or variable aperture. The modulating variations may be continuous analog variations or may be time modulated switching or gating. With reference to FIG. 2, the gas cluster ion beam apparatus 200, a pre-process metrology tool 420, and a post-process metrology tool 440 are configured to communicate with each other and with a host server 450. The pre-process metrology and post-process metrology tools 420, 440 may be film thickness measurement tools that use spectroscopic ellipsometry, scatterometry, interferometry, X-ray fluorescence, and four point probe techniques. In the embodiment shown in FIG. 2, the pre-process and post-process metrology tools 420, 440 are located ex-situ of the GCIB processing apparatus 200, which means that the measurement equipment is located outside the vacuum vessel 102 and separate from the GCIB processing apparatus 200. In another embodiment, the pre-process and post-process metrology tools 420, 440 are located in-situ and, to that end, may be contained within the vacuum vessel 102 (FIG. 1) to allow for in-vacuum measurements on the GCIB processing apparatus 200. In yet another embodiment, the pre-process and post-process metrology tools 420, 440 may be located in-situ outside the vacuum vessel 102 but still considered part of the GCIB processing apparatus 200. The pre-process and post-process metrology tools 420, 440 may be separate equipment, if well matched, or they may be the same metrology equipment. The pre-process metrology tool 420 and the post-process metrology tool 440 may communicate with the GCIB processing apparatus 200 by electrical signals communicated through a wired interface, such as a SEMI Equipment Communications Standard/Generic Equipment Model (SECS/GEM) wired interface. A SECS/GEM communication is a wired protocol between a host server 450 and the GCIB processing apparatus 200, pre-process metrology tool 420, and post-process metrology tool 440, as well as with other semiconductor manufacturing tools or equipment (not shown). SECS is a layer 6 protocol that describes the content of the messages while GEM is a higher layer application protocol that defines the messages themselves. Alternatively, the wired interface over which the electrical signals are communicated between the host server 450, GCIB processing apparatus 200, pre-process metrology tool 420, and post-process metrology tool 440 may be a registered jack (RJ) standardized physical interface such an eight-pin Ethernet (8P8C) or two-pin (RJ-11) connector, or a universal serial bus (USB) interface, or an RS-232 serial binary data connection. In one embodiment, a SECS/GEM communication is transferred between the gas cluster ion beam apparatus 200 and the pre-process metrology tool 420 over wired communication paths 415 and 455 via server 450. In another embodiment, the SECS/GEM communication is transferred between the GCIB processing apparatus 200 and the post-process metrology tool 440 through wired communication paths 425 and 455 via server 450. In a further embodiment, pre-process metrology tool 420 communicates with post-process metrology tool 440 through wired communication path 415 and 425 via server 450. Alternatively, the GCIB processing apparatus 200 and pre-process metrology tool 420, and the GCIB processing apparatus 200 and post-process metrology tool 440 may be coupled in communication using short-range wireless technology connections 410, 430, respectively, characterized by respective transceiver interfaces. In one embodiment, the wireless connections 410, 430 may comprise a short-range wireless technology connection to limit interference with other processing equipment, although the invention is not so limited as a long-range wireless connection may be used in an alternative embodiment. Short-range wireless technologies, such as Bluetooth wireless technology, may communicate data signals over a distance of up to 10 meters in a frequency range between 2.402 gigahertz (GHz) and 2.480 GHz. Bluetooth protocols are described in “Specification of the Bluetooth System: Core, Version 1.1,” published Feb. 22, 2001 by the Bluetooth Special Interest Group, Inc. Associated, as well as previous or subsequent, versions of the Bluetooth standard may also be supported by the wireless connections 410, 430. Alternatively, the short-range wireless technology, such as ultra-wideband (UWB), may communicate digital data over a wide spectrum of frequency bands ranging in a frequency range between 3.1 GHz and 10.6 GHz. Other examples of a short-range wireless technology includes certified wireless universal serial bus (USB), and communications defined by the Institute of Electrical Institute of Electrical and Electronic Engineers (IEEE) 802.11, Wireless Fidelity (Wi-Fi) and IEEE 802.16 Worldwide Interoperability for Microwave Access (WiMAX) suites of standards. IEEE 802.11b corresponds to IEEE Std. 802.11b-1999 entitled “Local and Metropolitan Area Networks, Part 11: Wireless LAN Medium Access Control (MAC) and Physical Layer (PHY) Specifications: Higher-Speed Physical Layer Extension in the 2.4 GHz Band,” approved Sep. 16, 1999 as well as related documents. IEEE 802.11g corresponds to IEEE Std. 802.11g-2003 entitled “Local and Metropolitan Area Networks, Part 11: Wireless LAN Medium Access Control (MAC) and Physical Layer (PHY) Specifications, Amendment 4: Further Higher Rate Extension in the 2.4 GHz Band,” approved Jun. 27, 2003 as well as related documents. In one embodiment, parametric data is transferred between the metrology tools 420, 440 and the GCIB processing apparatus 200 to correct for non-uniformities. According to this embodiment, film thickness map information is fed into the GCIB processing apparatus 200 as a parametric data file. Using a previously measured beam removal function and a previously measured relationship between etch rate and dose for a particular set of GCIB parameters (including GCIB energy and cluster species), a mathematical algorithm is then employed which takes the non-uniformity data, inverts beam spot etching pattern to fit the non-uniformity profile, and creates a beam-dose contour as a systematic error response to selectively remove surface material and thereby achieve a uniformly thick film. Many different approaches to the selection of mathematical algorithm may be successfully employed in this embodiment. In another embodiment, the beam-dose contour may selectively deposit additional material as a systematic error response on the surface to achieve a uniformly thick film. To a first approximation, a beam profile will be a Gaussian function for any cross-section slice of the beam 202 (FIG. 1) in cylindrical coordinates with the beam propagation axis as the Z-axis of the coordinate system. For the case of profiling by variations in the beam dwell time, the mathematical inversion and deconvolution that must be performed are simplified because the response function of the sample is linear with changes in dose. Hence, the beam removal function has essentially the same mathematical functional shape as the beam intensity profile. The beam dwell-time map, which directly determines the beam-scan pattern, must be implemented for each systematically varying film batch if angstrom-scale uniformity is desired. Once processed to GCIB specifications, the uniformity of the workpiece(s) may be examined either in-situ or ex-situ and the process finished or refined as appropriate. According to this embodiment, the GCIB processing apparatus 200 as shown in FIG. 2 has a facility for control of the beam-scan profile by direct feedback from the non-uniformity map data, as established either in-situ or ex-situ to the GCIB processing apparatus 200. Further, the in-situ measurement method may be preferred because it is the most time efficient method and it permits iteration without exposing the workpiece 210 to repeated vacuum/atmosphere cycles. FIG. 3 illustrates a film-thickness map of a wafer with an upper layer comprising a thin film or layer as measured by spectroscopic ellipsometry using a commercially available model UV-1280SE thin film measurement instrument manufactured by KLA-Tencor Corporation. As apparent from FIG. 3, the thickness of a thin film on a workpiece may be mapped as a function of position. The initial thickness non-uniformity of an upper film layer on workpiece 210 may be characterized ex-situ of the GCIB apparatus 200 by spectroscopic ellipsometry or other suitable conventional techniques. Such techniques can produce a point-by-point film thickness map that may be reduced to thickness contours (or similar) as shown in FIG. 3. Similarly, an in-situ uniformity-mapping instrument using spectroscopic ellipsometry or other suitable conventional film thickness mapping techniques may be incorporated within the GCIB apparatus 200 (FIG. 1) for guiding a profiling process. In either case, the non-uniformity measurements may be stored as a series of thickness points with precise workpiece positions by a standard computer. A film measurement method such as spectroscopic ellipsometry is used to map the thickness of only the top film layer, independent of variations in workpiece thickness, thickness of underlying films, or surface flatness. A flowchart is illustrated in FIG. 4 describing one embodiment of a method of creating a systematic error response based on un-modulated gas cluster ion beam data and averaged post-process parametric data. A systematic error response is used to correct for systematic non-uniformities of a process tool. Systematic non-uniformities are a result of inherent variations in a process tool such as a gas cluster ion beam tool. In block 500, a plurality of workpieces 210, such as wafers with a similarly situated upper film layer, are collected. In one embodiment, the plurality of workpieces 210 may be a cassette of wafers with a thin film uniformly applied on a top surface of each wafer using a deposition processing tool. In block 510, the plurality of workpieces 210 are processed using an un-modulated GCIB 202. In this embodiment, each workpiece 210 is processed on the GCIB processing apparatus 200 in an un-modulated manner, at constant beam intensity and at a constant scan speed so that all areas of an upper layer of the workpiece 210 receive an equivalent applied dose of cluster ions. In block 520, post-process parametric data is collected from the plurality of workpieces 210 using a post-process metrology tool. The post-process parametric data may be a collection of dielectric film thickness measurements, metal film thickness measurements, resist film thickness measurements, resistivity measurements, surface layer contamination measurements, surface roughness measurements, or other parametric data. In block 530, the post-parametric data is averaged. In block 540, systematic non-uniformities are identified based on the averaged post process parametric data. In block 550, a systematic error response is derived based at least in part on the averaged post-process parametric data. The systematic error response may be formulated at least in part on the etch rate or deposition rate of the gas cluster ion beam, non-uniformity map data for a workpiece to be processed, and a gas cluster ion beam flux profile wherein a density of ions across a diameter of the gas cluster ion beam is characterized. FIG. 5 is a flowchart describing one embodiment of a method of correcting non-uniformities based at least in part on a systematic error response. In block 600, incoming parametric data is collected from an upper layer of a workpiece 210. The incoming parametric data may be data collected using film thickness measurement tools using technologies such as spectroscopic ellipsometry, scatterometry, interferometry, X-ray fluorescence, and four point resistivity techniques. The incoming parametric data may be measured on a workpiece 210 to be processed by a gas cluster ion beam tool 200. In block 610, non-uniformities are identified in the incoming parametric data. In block 620, a GCIB 202 is directed toward a surface of the workpiece 210. In block 630, an applied dose from the GCIB 202 is modulated using a systematic error response to correct for systematic non-uniformities. Modulation represents an un-even application of dose to an upper layer either through a change in scan speed of the X-Y positioning table 204 as the GCIB 202 traverses the upper layer or through a change in intensity of the GCIB 202. Either method may be used individually or in combination to modulate an applied dose. An applied dose is a measure of the amount of material from a GCIB 202 impinging on a workpiece 210 over time. The material from the GCIB 202 may be incorporated into the workpiece in some form or it may escape the workpiece in the form of a gas or vapor. To change the scan speed, the X-Y controller 216 (FIG. 1) maneuvers the X-Y positioning table 204 to position every part of the workpiece 210 for processing according to predetermined velocities that result appropriate beam dwell times to etch away or to deposit material to the desired thicknesses to correct for systematic non-uniformities and to provide a film of uniform thickness. This is one embodiment for modulating an applied dose of the GCIB 202 (FIG. 1). Alternatively, a beam is scanned at constant velocity in a fixed pattern across the surface, but the GCIB intensity is modulated (often referred to as Z-axis modulation) to deliver an intentionally non-uniform dose to the sample. This is another embodiment for modulating an applied dose of the GCIB 202. The GCIB intensity may be modulated by any or a variety of methods, including for example but not limited to: by varying a gas flow from the GCIB source supply; by modulating the ionizer of the GCIB processing apparatus 200 either by varying a filament voltage VF, or by varying an anode voltage VA; by modulating the lens focus by varying lens voltages VL1 and/or VL2 in the GCIB processing apparatus 200; or by mechanically blocking a portion of the beam by means of a variable beam block, adjustable shutter, or variable aperture. The modulating variations may be continuous analog variations or time modulated switching or gating. A plurality of embodiments for correcting non-uniformities, including non-uniformities of an upper layer of an incoming workpiece and systematic non-uniformities caused by inherent variation in a process tool such as a gas cluster ion beam tool, using a gas cluster ion beam has been described. The foregoing description of the embodiments of the invention has been presented for the purposes of illustration and description. It is not intended to be exhaustive or to limit the invention to the precise forms disclosed. This description and the claims following include terms, such as left, right, top, bottom, over, under, upper, lower, first, second, etc. that are used for descriptive purposes only and are not to be construed as limiting. For example, terms designating relative vertical position refer to a situation where a device side (or active surface) of a substrate or integrated circuit is the “top” surface of that substrate; the substrate may actually be in any orientation so that a “top” side of a substrate may be lower than the “bottom” side in a standard terrestrial frame of reference and still fall within the meaning of the term “top.” The term “on” as used herein (including in the claims) does not indicate that a first layer “on” a second layer is directly on and in immediate contact with the second layer unless such is specifically stated; there may be a third layer or other structure between the first layer and the second layer on the first layer. The embodiments of a device or article described herein can be manufactured, used, or shipped in a number of positions and orientations. Persons skilled in the relevant art can appreciate that many modifications and variations are possible in light of the above teaching. Persons skilled in the art will recognize various equivalent combinations and substitutions for various components shown in the Figures. It is therefore intended that the scope of the invention be limited not by this detailed description, but rather by the claims appended hereto.
	abstract	Disclosed is an X-ray reflecting device and an X-ray reflecting element constituting the X-ray reflecting device capable of facilitating a reduction in weight and being prepared in a relatively simple manner. The X-ray reflecting element of the present invention comprises a body made of a solid silicon, and a plurality of slits formed in the body in such a manner as to penetrate from a front surface to a back surface of the body. Each of the slits has a wall surface serving as an X-ray reflecting surface. To allow the slits in the respective X-ray reflecting elements to be located in a given positional relationship with each other, the X-ray reflecting device of the present invention comprises a plural number of the X-ray reflecting elements, which are formed into a multilayered structure in such a manner or arranged side-by-side in a horizontal direction in such a manner as to allow the slits in the respective X-ray reflecting elements to be located in a given positional relationship with each other, or stacked on each other in a vertical direction to form a stacked structure in such a manner as to allow the slits in the respective X-ray reflecting elements to be located in a given positional relationship with each other. Further, the X-ray reflecting device may comprise a plural number of the stacked structures arranged side-by-side in a horizontal direction.
	claims	1. A pellet for use as a target in a device that produces thermonuclear fusion by inertial confinement, said pellet made by a process comprising the following steps: A. preparing monomers which contain exclusively carbon and tritium; B. polymerizing said monomers to produce a high polymer hydrocarbon containing only carbon and tritium, and C. forming a rigid, hollow pellet from said high polymer hydrocarbon by means of rotation that is spheroidal in shape and has a wall containing only carbon and tritium.
048266531	claims	1. In a fuel assembly for a boiling water reactor, said assembly being of the type having a vertical center line and being configured for control by control rods external to said assembly, said fuel assembly including a lower lattice element; a plurality of fuel rods supported by said lattice element; at least one fuel box surrounding a corresponding group of said fuel rods; a sleeve-like base supporting said lattice element and said at least one fuel box, said base having a downwardly directed first inlet opening for supplying a coolant flow to the surfaces of said fuel rods and a downwardly facing, annular end surface having an inner diameter defining said first inlet opening; and a bypass means supported by said assembly for conducting a by-pass flow of water along said fuel rods, the improvement wherein said bypass means comprises: first means for conducting water along the greater part of the length of said fuel rods without the water's contacting said fuel rods while the water is conducted therealong, said first mens having a first minimum flow area; and second means, positioned below said lattice element and connected in series with said first means, for conducting a portion of said flow to said first means, said second means having a second minimum flow area and a second inlet opening, said second inlet opening being positioned below a horizontal plane, said horizontal plane being located at a distance above the lowermost point of the fuel assembly, said distance being equal to said inner diameter of said first inlet opening, said first minimum flow area being greater than said second minimum flow area such that said second means constitutes a flow restriction for said bypass flow. first means for conducting water along the greater part of the length of said fuel rods without the water's contacting said fuel rods while the water is conducted therealong, said first means having a first minimum flow area; and second means, positioned below said lattice element and connected in series with said first means, for conducting a portion of said flow to said first means, said second means having a second minimum flow area and a second inlet opening, said sleeve-like base causing said throttled stream to acquire at a level above said first inlet opening an increased cross-section with only vertical stream lines and said second inlet opening being positioned below said level, said first minimum flow area being greater than said second minimum flow area such that said second means constitutes a flow restriction for said bypass flow. 2. The improvement according to claim 1, wherein said second means comprises an upwardly directed supply tube, an ejector pump positioned within said sleeve-like base and connected to receive water from said supply tube, at least one suction tube extending from the exterior of said sleeve-like base to deliver water to said ejector pump, an outlet nozzle connected to said ejector pump to deliver water therefrom, and means connected to said outlet nozzle for distributing water to said first means for conducting water along said fuel rods. 3. The improvement according to claim 1, wherein there are a plurality of said fuel boxes supported by said base, said fuel boxes being spaced from each other to define a water channel thereamong; and said second means comprises an upwardly directed supply tube having an upper end opening into said water channel. 4. The improvement according to claim 1, wherein said second means comprises an upwardly directed supply tube having an upper end, third means connected to said upper end for distributing water to said first means, said third means defining an interior cavity hydraulically communicating with said first means, and hydraulic connector means, extending from said cavity to the exterior of said fuel assembly, for introducing a flow of water into said cavity simultaneously with the flow from said supply tube. 5. The improvement according to claim 4, wherein said first means comprises a plurality of vertical water tubes arranged in two rows which intersect at said center line, the lower ends of said water tubes being supported by said third means and opening into said interior cavity; said third means comprises a cruciform hollow body defining said cavity, said cruciform hollow body being supported by said sleevelike base and having a central inlet opening connected to said supply tube; and said hydraulic connector means comprises tubes extending from said cavity out of said cruciform hollow body to the exterior of said fuel assembly. 6. The improvement according to claim 1, wherein said second means comprises a supply tube arranged to conduct water upwardly from said inlet opening and laterally through the side wall of said sleeve-like base; third means, defining an interior cavity hydraulically communicating with said first means, for distributing water to said first means; and hydraulic connector means extending from said cavity to the exterior of said fuel assembly, for introducing a flow of water into said cavity in response to flow through said supply tube. 7. The improvement according to claim 6, wherein said hydraulic connector means comprises tubes extending from said cavity to the exterior of said fuel assembly. 8. The improvement according to claim 1, wherein said flow restriction provides a magnitude of said bypass flow for optimum distribution of said water conducted by said first means at full reactor power, and a magnitude of said bypass flow sufficient to prevent significant boiling of said water conducted by said first means upon maximum reduction of said full reactor power. 9. The improvement according to claim 1, wherein said coolant flow comprises a throttled stream of water, and said second inlet opening is below a level at which said throttled stream of water acquires an increased crosssection with only vertical streamlines. 10. The improvement according to claim 1, wherein a water throttling means is provided below said first inlet opening, and wherein said fuel assembly is arranged with its lowermost point at a vertical distance above said throttling opening of at the most substantially 20 mm. 11. The improvement according to claim 10, wherein said vertical distance is less than 10 mm. 12. The improvement of claim 2, wherein at least substantially one-fourth of the water conducted by said first means is delivered to said ejector pump by said at least one suction tube. 13. In a fuel assembly for a boiling water reactor, said assembly being of the type having a vertical center line and being configured for control by control rods external to said assembly, said fuel assembly including a lower lattice element; a plurality of fuel rods supported by said lattice element; at least one fuel box surrounding a corresponding group of said fuel rods; a sleeve-like base supporting said lattice element and said at least one fuel box, said base having a downwardly directed first inlet opening for supplying a coolant flow to the surfaces of said fuel rods and a downwardly facing, annular end surface having an inner diameter defining said first inlet opening, said coolant flow first passing through a throttling opening below said first inlet opening to provide a throttled stream of water; and a bypass means supported by said assembly for conducting a by-pass flow of water along said fuel rods, the improvement wherein said bypass means comprises: 14. The improvement according to claim 13, wherein said second inlet opening is below a horizontal plane located at a distance above said throttling opening equal to said inner diameter of said first inlet opening plus not more than 20 mm. 15. The improvement according to claim 13, wherein said second inlet opening is below a horizontal plane located at a distance above said throttling opening equal to said inner diameter of said first inlet opening plus less than 10 mm.
	claims	1. A process for the manufacture of a dense material containing uranium dioxide UO2, the process comprising:introducing a plurality of powders each including uranium dioxide UO2, into a vibrating grinder, at least two of the powders provided by, respectively, a wet route and a dry route, the at least two powders having different crystallite agglomeration and aggregation degrees, each of the powders having a specific surface area within about 10 m2/g of each other powder;shaking the powders with the vibrating grinder so as to form a particulate material lacking agglomerates and aggregates, the shaking including breaking up agglomerates and aggregates in the powders without at the same time breaking up crystallites in the powders, the shaking destroying substantially all agglomerates and aggregates in the powders;introducing the particulate material into a mould;applying a shaping stress to the particulate material so as to obtain a compacted material; andsintering the compacted material so as to obtain the dense material. 2. The process according to claim 1, wherein at least one of the powders contains at least one oxide chosen from the group consisting of uranium oxide U3O8, uranium oxide U3O7, plutonium oxide PuO2, and thorium oxide ThO2. 3. The process according to claim 1, wherein at least one of the powders contains at least one additive chosen from the group consisting of gadolinium oxide Gd2O3, erbium oxide Er2O3, a pore forming substance, a lubricant, and a sintering promoter. 4. The process according to claim, 1 wherein the shaking is carried out over at least a duration so as to form a particulate material having a substantially constant sinterability, compressibility and sinterability of the particulate material having values independent from the amount of agglomerates and/or aggregates contained in each of the powders. 5. The process according to claims 1, wherein the shaking is carried out by the grinder so as to increase the compressibility of the particulate material up to a substantially constant value, while sinterability remains substantially constant. 6. The process according to claim 1, wherein the applying the shaping stress is carried out under a stress ranging from about 200 to 1200 MPa. 7. The process according to claim 1, wherein each powder has a specific surface area within about 2 to 5 m2/g of each other powder.
	summary
059784300	description	DETAILED DESCRIPTION OF THE INVENTION Referring now to the drawings particularly to FIG. 1, there is illustrated a schematic representation of a nuclear fuel bundle generally designated 10 comprised of a lower tie plate 12, a pair of water rods 14 secured at their lower ends to the tie plate 12, spacers 16 secured to the water rods 14 at spaced positions therealong and fuel rods 18 received through openings in the spacers 16 and secured to the lower tie plate 12. The water rods 14 are each comprised of essentially two sections, a lower central tube 20 and a tie bar assembly 22. The central tube 20 and tie bar assembly 22 of each water rod are secured to one another in a length adjustable manner by a structure which does not form part of the present invention. Suffice it to say that the lengths of the water rods 14 are adjustable in the axial direction within a limited range. As indicated previously, it is important in the present invention to ensure that the lengths of the water rods 14 from the upper surface of the lower tie plate 12 to the underside of seats 24 at the upper ends of the tie rod assemblies 22 are within designed lengths. Referring to FIGS. 2-5, the length gauge of the present invention includes an indicator location block 26, a digital indicator gauge 28, a standards rod 34 and an indicator set block 37. The digital indicator gauge 28 has a digital readout 30 and a movable extension pin 32. Location block 26 includes a pair of generally semi-cylindrical openings or recesses 38 open to opposite sides of body 26. The upper end of body 26 has a boss 39 with a central aperture 40 for receiving a boss 42 on gauge 28. The gauge 28 and extension pin 32 may be suitably secured to the boss 39 at location 36. Block 26 has a first reference surface 43 for engaging in use the seat 24 (in FIG. 1) of a tie rod assembly. The tip of the movable extension pin 32 constitutes a second reference surface 44. The standards rod 34 comprises an elongated gauge lab length certified cylindrical rod. The certified length of rod 34 from the face of the lower tie plate 12 on FIG. 2 to the top, i.e., flat surface 45, of the standards rod is an accurate specified length which when combined with the length between the first and second reference surfaces 43 and 44, respectively, on the indicator location block and gauge, with the gauge zeroed out after calibration, is the same as the specified length of the water rod from the face of the lower tie plate 12 to the seat 24. The standards rod 34 has a reduced diameter portion 46 at its lower end for reception in an opening not shown in the upper surface of the lower tie plate 12, such that the shoulder 48 abuts the tie plate surface. The opposite end of the standards rod 34 includes a hardened bearing surface 45 for receiving the tip 44 of the movable extension pin 32 of the gauge. Referring to FIGS. 3 and 4, the indicator set block 37 comprises a flat plate 50 having a pair of cylindrical projections 52 substantially straddling a reduced length cylindrical projection 54. The upper surfaces of the projection 52 and 54 constitute first and second datum surfaces, 56 and 64 respectively, of known distance between such surfaces used to provide the gauge with a predetermined indication; for example, as a zero position. To calibrate the gauge, one of the projections 52 is disposed in a recess 38 of the indicator location block 26 with its datum surface 56 bearing against a third reference surface 58 formed by a shoulder in part defining the recess 38. With the first datum surface 56 on the indicator set block 37 engaged against the third reference surface 58 on the indicator location block 26, and the second reference surface comprising the tip 44 of movable extension pin 32 bearing against the second datum surface 64 on the indicator set block, the gauge 28 can be zeroed out, i.e., manually adjusted to provide a zero reading. Prior to assembling the fuel bundle, the water rods 14 in FIG. 1 are secured to the tie plate 12 in a known manner, the assembly being positioned in a horizontal orientation. The length gauge is then employed to measure the lengths of the water rods so that adjustments of the axial length of the water rods can be made to meet the design lengths. To accomplish this, the reduced end 46 of the standards rod 34 in FIG. 2 is inserted into an opening of the tie plate 12 until shoulder 48 abuts the top surface of the tie plate. The indicator location block 26 is applied below the seat 24, i.e., a datum surface of one of the water rods, such that the arcuate recess 38 receives the reduced diameter upper end of the tie bar assembly 22 as illustrated in FIGS. 1 and 2. In that configuration the second reference surface, i.e., the tip 44 of the movable extension pin 32, bears against the hardened upper surface 45 of the standards rod 34. With the first reference surface 43 bearing against the seat 24 and the second reference surface 44 of movable extension pin 32 bearing against the upper end 45 of the standards rod 34, the gauge 28 will indicate any deviation of the length of the water rod from a designed length corresponding to the length of the standards rod plus the predetermined distance between the second reference surface 44 and the seat 24. Should a deviation be indicated by the digital readout 30 outside of required tolerances, the water rods must be adjusted in axial length to more precisely conform to the design length. Once the water rods have been adjusted in length, the gauge assembly (26 and 28) and the standards rod 34 are removed from the partial assembly. Next, the bundle 10 is assembled starting with the installation of the spacers 16 to the water rods 14. It will be appreciated that the spacers 16 are secured to the water rods 14 at axially spaced locations along the length of the fuel bundle 10 with each spacer having openings as conventional for receiving the water rods and the fuel rods 18. After securement of the spacers to the water rods, the standards rod is installed in the fuel bundle. To accomplish this, and referring to FIG. 6, a shaped end, for example a bullet nose 64, is installed on the reduced diameter end 46 of the standards rod 34. The standards rod is then inserted from the top of the fuel bundle through openings in the spacers toward the lower tie plate 12. The shaped end 64 facilitates insertion of the standards rod through the openings of the spacers. When the end of the standards rod extends in the region between the lowermost spacer and the lower tie plate 12, the shaped end 64 is removed and the reduced end portion 46 is inserted into the opening of the tie plate similarly as before, as illustrated in FIG. 7. With the standards rod in place, the fuel rods 18 are then inserted through the openings in the spacers and secured to the lower tie plate 12. If desired, an additional verification of the lengths of the water rods can be made at this time, as described in the previous paragraph. Lastly, a suitable lifting tool, not shown, is installed on the bundle to move the bundle from the horizontal assembly position to a vertical inspection stand, not shown. With the fuel bundle in a vertical orientation, and with the lifting tool removed, the gauge and indicator location block are applied to each water rod and measurements are taken similarly as previously described. Thus, any deviation of the length of the water rod from its designed length will be indicated by the gauge 28. Once the measurements are made and the water rods are verified to lie within the designed length, the gauge is removed and the standards rod is withdrawn axially from the tie plate and spacers. It will be appreciated that the standards rod is inserted in one of the openings of each spacer which is otherwise occupied by a fuel rod. Once the standards rod has been removed, a full length or part length fuel rod is inserted in the position previously occupied by the standards rod. While the invention has been described in connection with what is presently considered to be the most practical and preferred embodiment, it is to be understood that the invention is not to be limited to the disclosed embodiment, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims.
046408118	summary	FIELD OF THE INVENTION The present invention relates to a linear motion device for vertically moving a control rod in the core of a nuclear reactor. BACKGROUND In a Pressurized Water Reactor (PWR), a certain number of linear motion devices are used for controlling the power of the reactor. Each device comprises a vertical control shaft connected to the upper end of the control rod and moved by electromagnetically actuated pawls. Such a device described, for example in the U.S. Pat. No. 3,158,766, allows a step by step movement of the control rod consisting of a cluster of neutron absorbing rods, in a fuel assembly of the core along the vertical direction. Each device is arranged in a sealed vessel in communication with the inside of the reactor vessel and mounted on its cover. A motion device, called control rod mechanism comprises two sets of pawls (or dogs), one set being mounted on a first support in the sealed vessel and the other being mounted on a second support movable in the vertical direction of the sealed vessel allowing the transfer of the control shaft and the control rod, The control shaft is provided with grooves equally spaced in the vertical direction and each dog of each set comprises a tooth engageable with play in a groove by rocking of the pawl under the effect of an electromagnetic actuation or impulse. Each dog is articulated on its support through an upper horizontal axle and on a connecting rod through a lower horizontal axle. The connecting rod is articulated at its end opposite to the pawl on a part of the electromagnetic actuation means. The rocking of the pawl is obtained through the connecting element. The support of the transfer pawl is also actuated by an electromagnetic means for is displacement along a vertical distance equalling that separating two successive grooves of the control shaft and corresponding to one step of displacement. The holding pawls are actuated to engage their teeth in a groove of the control shaft at the end of a displacement to allow the return of the transfer pawls to their original position. The transfer pawls are actuated to engage the control shaft before effecting a displacement of the control shaft. After the engagements of one set of pawls, a slight upward displacement of their support allows a disengagement of the other set of pawls without any friction. Generally speaking, every movement of the pawls discribed in U.S. Pat. No. 3,158,766 is made without friction by means of the play between the teeth of the pawls and the grooves of the control shaft. Thus, the wear on the motion device is greatly reduced. However, when the nuclear reactor is used with frequent displacements of the control rods, the wear on the dogs may become excessive after a certain period of operation and what is worse, a tooth may break and be separated from the pawl. The physical phenomena responsible for the wear and breakage of the pawls in the control rod mechanisms are not accurately known at this time, except the fact that shock and vibrations are important factors in such discontinuous mechanisms. It is necessary for the mechanisms to be sufficiently reliable to bear at least 10 million stops without breakage and without an excessive wear, under the conditions of use involving large stresses such as in the high power reactors built in recent years (1300 MW), which is not possible with the mechanisms of the prior art. It has been previously proposed to replace mechanisms with pawls by other types of mechanisms such as screw and nut or electromagnetic coupling mechanisms. However the design of the reactor would have to be partly modified and adapted to the new type of mechanism. SUMMARY OF THE INVENTION It is therefore an object of the invention to provide a linear motion device for vertically moving a control rod in the core of a nuclear reactor comprising, in a sealed vessel in communication with the inside of the reactor vessel, a vertical control shaft provided with grooves equally spaced in the vertical direction, connected to the control rod at its lower end and two sets of pawls, on set being mounted on a first support in the sealed vessel for holding the control shaft and the other on a second support movable in the vertical direction in the sealed vessel for moving the control shaft, each pawl of each set being articulated on the corresponding support through a first horizontal axle, thus permitting the rocking of said pawl to engage or disengage toothed means formed by the pawl in a grooved means of the control shaft, by a displacement of the actuation means, the said mechanism permitting a very long life span of operation without excessive wear or breakage of the pawls, even in the case of frequent use and heavy stresses on the pawls. For this purpose, each of said pawls comprises two engaging teeth, one upper tooth and one lower tooth, the tips of which are separated by a vertical distance equalling that separating two successive grooves of the control shaft and said lower axle of each of said pawls is centered at a location on said pawl situated under and at a vertical distance from the tip of said upper tooth between 1/6 and 1/4 of the vertical distance separating the two tips of said engaging teeth on said pawl. In order to make the invention more fully understood, an embodiment of a linear motion device for moving a control rod in the core of a Pressurized Water Reactor will now be described with reference to the accompanying drawings, purely by way of non limiting example.
	abstract	A heat pipe reactor may include a reactor core and one or more heat exchangers positioned on one or both sides of the reactor core. The heat pipe reactor may also include a plurality of heat pipes extending from the reactor core and out through the one or more heat exchangers. The reactor core may be composed of a plurality of monolithic blocks.
	summary
	claims	1. A basket comprising:a square pipe assembly having a plurality of square pipes assembled in a staggered arrangement, wherein each of the plurality of square pipes comprises:four walls and four corners arranged so as to form a square cross section, the four walls including neutron-absorbing material and having a thickness capable of preventing spent fuel inserted therein from reaching criticality and capable of ensuring a predetermined strength at a time of falling down, wherein a cell for housing a spent fuel aggregate is formed in a space defined by walls of the square pipes;a connecting section formed on each of the four corners, at which diagonally adjacent square pipes are contacted with each other, wherein the connecting section is formed into a terrace shape having a plurality of steps, and the steps of the terrace shape are formed to butt against the steps of adjacent square pipes when the square pipes are assembled in a staggered arrangement; anda square pipe receiver placed between adjacent square pipes located on the outermost circumference of the square pipe assembly. 2. The basket according to claim 1, further comprising:a fastener which fastens the outermost square pipes of the square pipe assembly; anda spacer block formed to fit to the inner shape of the spent fuel container. 3. A spent fuel housing container comprising:a basket according to claim 1 formed into a lattice shape as a whole by combining a plurality of square pipes with each other in a staggered arrangement and inserted to the spent fuel housing container to be fitted to cavity inner shape of the spent fuel housing container so that a spent fuel aggregate is housed inside the lattice. 4. A basket comprising:a square pipe assembly having a plurality of square pipes assembled in a staggered arrangement, wherein each of the plurality of square pipes comprises:four walls and four corners arranged so as to form a square cross section, the four walls including neutron-absorbing material and having a thickness capable of preventing spent fuel inserted therein from reaching criticality and capable of ensuring a predetermined strength at a time of falling down, wherein a cell for housing a spent fuel aggregate is formed in a space defined by walls of the square pipes;a connecting section formed on each of the four corners, at which diagonally adjacent square pipes are contacted with each other, wherein the connecting section is formed into a terrace shape having at least one step, and the step of the terrace shape is formed to butt against the step of adjacent square pipes when the square pipes are assembled in a staggered arrangement; anda square pipe receiver placed between adjacent square pipes located on the outermost circumference of the square pipe assembly. 5. A spent fuel housing container comprising:a basket according to claim 4 formed into a lattice shape as a whole by combining a plurality of square pipes with each other in a staggered arrangement and inserted to the spent fuel housing container to be fitted to cavity inner shape of the spent fuel housing container so that a spent fuel aggregate is housed inside the lattice. 6. The spent fuel housing container according to claim 5, including portions where no side faces are present along the outer circumference of the basket, wherein said square pipe receiver is provided at the portions where there are no side faces along the outer circumference of the basket. 7. The spent fuel housing container according to claim 6, wherein said square pipe receiver is comprised by a convex portion of the spent fuel housing container.
	claims	1. A charge particle beam apparatus comprising:a charged particle beam lens barrel configured to irradiate a charged particle beam;a gas nozzle configured to feed gas near a position on a sample at which the charged particle beam is irradiated; anda gas supply amount adjusting device configured to be periodically operated to control a flow rate of gas between a gas supply source and the gas nozzle to intermittently feed the gas to the gas nozzle side. 2. The charged particle beam apparatus according to claim 1, comprising a gas diffusion mechanism provided between the gas nozzle and the gas supply amount adjusting device,wherein a diameter of the gas diffusion mechanism is larger than diameters of a gas pipe arrangement connected to the gas supply amount adjusting device and the gas nozzle. 3. The charged particle beam apparatus according to claim 1, wherein a chemical change is caused on an irradiation surface by a reaction of gas with a charged particle beam, and a structural change is generated at that location. 4. The charged particle beam apparatus according to claim 1, wherein an electrical charge on a charged particle beam irradiation surface is neutralized by gas. 5. The charged particle beam apparatus according to claim 1, wherein the charged particle beam is an ion beam. 6. The charged particle beam apparatus according to claim 1, wherein the charged particle beam is an electron beam. 7. A charged particle beam apparatus comprising:a charged particle beam lens barrel that irradiates a charged particle beam onto a sample;a gas nozzle that feeds gas toward a region of the sample irradiated by the charge particle beam;a control device configured to control the flow rate of gas fed from a gas supply source by intermittently flowing the gas along a flow path to the gas nozzle; anda gas diffusion mechanism disposed in the flow path between the gas nozzle and the control device for diffusing and uniformizing the intermittent gas flow before it reaches the gas nozzle. 8. A charged particle beam apparatus according to claim 7; wherein the control device is configured to intermittently feed the gas in repeated cycles in which, in each cycle, the time during which gas is fed to the gas diffusion mechanism is less than the time during which gas is not fed from the diffusion mechanism. 9. A charged particle beam apparatus according to claim 7; including gas pipes connecting the gas nozzle to the gas diffusion mechanism and connecting the gas diffusion mechanism to the control device, the gas diffusion mechanism having a larger diameter than that of the gas pipes. 10. A charged particle beam apparatus according to claim 7; wherein the length of the gas diffusion mechanism is at least ten times longer than the distance through which the gas flows in one cycle. 11. A charged particle beam apparatus according to claim 7; wherein the charged particle beam is an ion beam. 12. A charged particle beam apparatus according to claim 7; wherein the charged particle beam is an electron beam. 13. A charged particle beam apparatus according to claim 1; wherein the gas supply amount adjusting device is periodically operated to obtain a gas flow rate on the order of 0.02 ml/min. 14. A charged particle beam apparatus according to claim 1; wherein the gas supply amount adjusting device has a minimum gas supply amount when operating continuously and is periodically operated to obtain a gas supply amount as small as one-tenth the minimum gas supply amount. 15. A charged particle beam apparatus according to claim 1; wherein the periodic operation of the gas supply amount adjusting device comprises ON/OFF cycles of operation in which the OFF period is multiple times longer than the ON period. 16. A charged particle beam apparatus according to claim 8; wherein in each cycle, the time during which gas is not fed is multiple times longer than the time during which gas is fed. 17. A charged particle beam apparatus according to claim 7; wherein the control device has a minimum flow rate when operating continuously and is configured to operate periodically to reduce the flow rate to a value below the minimum flow rate. 18. A charged particle beam apparatus according to claim 17; wherein the control device is configured to operate periodically to obtain a flow rate as low as 0.02 ml/min. 19. A charged particle beam apparatus comprising:a charged particle beam lens barrel that irradiates a charged particle beam onto a sample;a gas nozzle that feeds gas toward a region of the sample irradiated by the charged particle beam;control means for controlling the flow rate of gas fed from a gas supply source by intermittently flowing the gas along a flow path to the gas nozzle; anda gas diffusion mechanism disposed in the flow path between the gas nozzle and the control device for diffusing and uniformizing the intermittent gas flow before it reaches the gas nozzle. 20. A charged particle beam apparatus according to claim 19; wherein the control means is configured to intermittently feed the gas in repeated cycles in which, in each cycle, the time during which gas is fed to the gas diffusion mechanism is less than the time during which gas is not fed from the gas diffusion mechanism.
047708476	summary	BACKGROUND OF THE INVENTION Fuel assemblies in certain boiling water nuclear reactors contain fuel rods and water rods. Typically, cladding containers of both types of rods are made of zirconium alloys such as Zircaloy-2 and Zircaloy-4. The fuel rods contain fuel material, such as uranium dioxide pellets whereas the water rods have water flowing through them. Under the conditions prevailing in an operating nuclear reactor, zirconium alloys exhibit a certain amount of irradiation growth. The amount of irradiation growth depends on many factors, including the alloy material, the temperature, neutron energy, and the neutron flux. Mechanical interaction between fuel pellets and the cladding container introduces a growth component in the fuel rods not present in water rods. This introduces an irradiation growth differential between fuel rods and water rods. As the radiation exposure increases, pellet-cladding mechanical interaction (PCMI) increases and as a result the difference in irradiation growth between fuel rods and water rods increases. In such a nuclear fuel assembly the fuel rods are positioned lengthwise between upper and lower tie plates in which the ends of the fuel rods are secured, for example by means of end plugs having studs fitted into holes in the tie plates. Water rods are positioned lengthwise between the tie plates and are likewise secured by means such as end plugs with studs fitted into corresponding holes in the tie plate. There is typically an axial irradiation growth differential among fuel rods and water rods. A small axial growth differential is accommodated, for example, by incorporation of expansion springs between the upper tie plate and the upper end of the fuel rod cladding. The expansion springs are disposed around a portion of the end plug studs. The distance between the upper and lower tie plates increases according to the axial irradiation growth of the fuel rods. There is concern that the difference in axial irradiation growth between fuel rods and water rods may become too great to be accommodated by the expansion springs and may cause one or more end plug studs of water rods to release from the upper or lower tie plate. SUMMARY OF THE INVENTION In accordance with the invention there is a process for providing different irradiation growth rates between first and second cladding tubes made of anisotropic metals such as zirconium alloys. The process comprises regulating the crystallographic structure of each cladding tube wherein low irradiation growth occurs when the crystalline structure of the metal is in a low energy state or configuration and larger irradiation growth occurs when crystalline structure of the metal is in a higher energy state. A preferred means for regulating the crystalline structure of the anisotropic metal in each cladding tube comprises tailoring the tube fabrication schedule to achieve the desired crystallographic structure. More specifically, the process comprises forming a first cladding tube of an anisotropic metal, preferably a zirconium alloy by a first fabrication schedule. The final dimensions are achieved by a selected cold-work reduction, which imparts a high energy state to the crystalline structure of the metal. The final cold-work reduction is followed by heat treating the first cladding tube at a selected temperature and for a selected time to allow the metal to at least partialy recrystallize to a lower energy state. A second cladding tube of the same metal as the first cladding tube is formed by a second fabrication schedule which includes a selected cold-work reduction to achieve given final dimensions which may be approximately the same as the first cladding tube. The second cladding tube may then be heat treated at a selected temperature and for a selected time sufficient to induce less recrystallization than the first cladding tube to thereby retain a higher energy state of the crystalline structure than that of the first tube. The second cladding tube will display greater irradiation growth than the first cladding tube due to its higher energy state. The energy state of the crystallographic structure of the metal is thus regulated by controlling the size of the final cold-work reduction in the tube fabrication schedule, the temperature of the heat treatment following the final cold-work reduction and the length of the heat treatment. A preferred embodiment of this invention comprises controlling the irradiation growth of fuel rods and water rods in a water cooled reactor. Proper tailoring of the cladding tube fabrication schedules enables water rods, which typically undergo less irradiation growth than fuel rods under operating conditions of a boiling water reactor, to exhibit substantially equivalent irradiation growth as fuel rods.
048881503	description	The control rod shown in FIGS. 1 and 2 is mainly made of stainless steel and consists of an absorber part 12 supported by a vertically arranged coupling bar 11. The absorber part 12 consists of four absorber plates 13-16, which form a rectilinear cross, the centre line of which coincides with that of the coupling bar 11. The absorber plates are provided at the centre with recesses 17 and projections 27 and are welded to each other at the projections 27 and at a supporting block 28 with square cross-section arranged adjacent the projections 27. The plates, which have a thickness of 8 mm, are provided with a large number of bored channels 18 (18a-c) with a diameter of 6 mm and a depth of 100 mm. In order to make FIG. 1 more clear, the channels therein are drawn without the distances which in reality exist between them, as will also be clear from the following. The uppermost channels 18a, which are most subjected to radiation, are filled with rods of hafnium metal of the same length as the channels. Hafnium does not swell upon irradiation. The channels 18b arranged below channels 18a are filled with powdered boron carbide 20 with the exception of the outermost part which is filled with a circular-cylindrical body 21 of hafnium metal. The body 21 has a diameter which is somewhat smaller than the diameter of the channel. The channels 18c arranged at the bottom, where the irradiation is lowest, are completely filled with powdered boron carbide 20 of the kind mentioned above. The distance between the envelope surfaces on two adjacent channels of types 18a and 18c is at all places 2 mm. For each channel 18b, the distance to the adjacent channel on one of its sides is 0.4 mm (the distance designated a in FIG. 3), and the distance to the adjacent channel on the other side is 3.6 mm (the distance designated b in FIG. 3). The edge of each absorber plate facing away from the centre line of the rod is provided with a slot 22, in which a bar 23 is arranged. At least in its upper part the bar 23 is suitably of hafnium metal. It may also in its entirety, and at least in the lower part, be of, for example, stainless steel. The application of the bar 23 in the slot 22 and the sealing of the slot 22 are described in greater detail below with reference to an explanation of FIGS. 3-5. For guiding of the control rod in the relatively narrow gaps between the fuel boxes of the reactor, the upper part of the control rod is provided with guide pads 24 of Inconel. In addition, it is provided with a lifting handle 25 for handling the rod during insertion and replacement. At its lower part, the rod is provided with a coupling head 26 over which the rod can be connected to a drive mechanism (control device). FIG. 3 shows a number of channels 18b in the absorber plate 15 as well as the slot 22, which in the exemplified case has a width of 5.5 mm in its entire length. From the information given above concerning the thickness of the absorber plate 15, the diameter of the channels 18, and the distances a and b between different channels of type 18b, it is clear that for each channel 18b the distance (a) from its envelope surface to the envelope of an adjacent channel is smaller than the distance (c) from its envelope surface to the surface of the absorber blade. As will be more clear from FIGS. 4 and 5, the absorber plate 15 has an edge portion 30, which comprises a gas-tight edge 31, a longitudinal space 32 located inside that edge, and a longitudinal bar 23 located in the slot 22 and having the same width as the slot 22 and being provided with a longitudinal indentation 33 which constitutes a limiting wall for the space 32. The bar 23 does not completely cover the orifices 34 of the channels in the bottom of the slot 22 as its width is smaller than the diameter of the channels, which, as mentioned above, amounts to 6 mm in the exemplified case. This leads to the creation of a gap 35 between the bar 23 and the side walls of the slot 22, where the channels are located, by the side walls of the slot 22 making contact with the bar 23 in the portions 38 (FIG. 3) where no channels are bored and, consequently, the side walls of the slot 22 are thicker. The channels 18b, as well as the channels 18a and 18c, are in open communication with the space 32 via the gaps 35, so that gas formed upon irradiation of the powdered absorber material may flow between the different channels in the plate 15 and pressure equilization be achieved. The gap 36 between the body 21 of hafnium metal and the inner wall of the channel 18b allows gas flow but prevents or radically counteracts a simultaneous transport of absorber material from the channel. In this way, a transport of absorber material from an undamaged channel to a channel exhibiting a crack, and through there to the surroundings of the control rod, is prevented or counteracted. Thus, the gap 36 (the distance between the inner wall of the channel 18b and the surface of body 21) is thinner than the gap 35 (the distance between the inner wall of channel 31 and the surface of bar 23 at the side of space 32) and, in addition, longer. The gap 36 is also thinner than the cross-section of the space 32, i.e. thinner than transport paths for the gas which are located in the edge portion. In the manufacture of an absorber plate, the channels 18 are bored from an edge on a plane-parallel plate. Thereafter, the slot 22 is milled out. After filling the channels with absorber material, the bar 23 is applied in the slot 22 and the side walls of the slot 22, which are straight from the beginning, are pressed against the bar 23 and welded together at the end surfaces while forming a gas-tight edge 31. The welded joint is designated 37. Also at its ends at the top and bottom of the absorber plate, the slot 22 is sealed by welded joints. The invention has been described in detail in connection with the use of boron carbide as swelling absorber material. The invention is also applicable to the use of other known absorber materials such as europium in the form of oxide or another compound. It is, of course, also applicable to control rods where no bodies 21 of hafnium are arranged in the outermost portions of certain absorber plates.
047568771	claims	1. A nuclear reactor having a pressure vessel and a core barrel having a bottom core support plate situated within the pressure vessel, the core support plate engaged about the periphery thereof by a plurality of engagement means, which have a recess therein for engagement with a key, wherein: said core support plate has a plurality of apertures therethrough one of which communicates with each recess of said engagement means, wherein a circular wall is provided in said core support plate about said apertures; and a key insertable into and positioned in said apertures in a secure relationship, said key having a lower section thereof of a rectangular cross-section which extends into said recess of said engagement means, said apertures of said core support plate being alignable with said engagement means of said pressure vessel and said keys being securable in said apertures of said core support plate and said recess of said engagement means from above said core support plate. said core support plate has a plurality of apertures therethrough one of which communicates with each recess of said engagement means and an upper surface with a countersunk shoulder about each aperture; a key insertable into and positioned in said apertures in a secure relationship, said key having a flange which mates with said shoulder to form a substantially horizontal upper surface on said core support plate after placement of said key, and a lower section thereof which extends into said recess of said engagement means, and fastening means to secure the flange of said key to said shoulder; said apertures of said core support plate being alignable with said engagement means of said pressure vessel and said keys being securable in said apertures of said core support plate and said recess of said engagement means from above said core support plate. providing a pressure vessel having a plurality of engagement means about the lower inner wall thereof, the engagement means having a recess in the upper surface thereof; providing said core support plate with a plurality of apertures therethrough, with circular walls about said apertures, one said aperture provided for alignment with the recess of each said engagement means; inserting said core barrel into the pressure vessel and initially aligning the apertures of the core support plate with the recesses of said engagement means; inserting a key having a cylindrical section slidable into said aperture and a non-cylindrical lower section into each aperture of the core support plate such that a lower section thereof extends into said recess; altering the lower section of the key to provide mating relationship between said lower section thereof and the recess; securing said keys in said core support plate with the lower section thereof mating with the recess of said engagements, and said apertures of said core support plate being aligned with said engagement means of said pressure vessel and said keys secured in said apertures of said core support plate and said recess of said engagement means from above said core support plate. 2. A nuclear reactor having a pressure vessel and a core barrel having a bottom core support plate situated within the pressure vessel, the support plate engaged about the periphery thereof by a plurality of engagement means, which have a recess therein for engagement with a key, wherein: 3. A nuclear reactor as defined in claim 2, wherein the fastening means comprises welding of the flange of said key to the core support plate. 4. A nuclear reactor as defined in claim 2, wherein bores are provided about the periphery of said flange of said key and complementary bores are provided in the shoulder of said core support plate and said fastening means are bolts engaged in said bores. 5. A method of aligning a core support barrel having a core support plate at the bottom thereof in a nuclear reactor pressure vessel comprising: 6. The method of claim 5 wherein a core support plate is provided having a countersunk portion forming a shoulder about the apertures and said keys having a flange seatable on said shoulder whereby the top of the key is flush with the top surface of the core support plate upon insertion into said aperture. 7. The method of claim 6 wherein said securing is by welding of the flange of said key to the core support plate. 8. The method of claim 6 wherein said securing is by bolting the flange of said key to said core support plate.
058833943	description	In the various figures of the drawing, identical features will be indicated with the same reference numerals, and similar features in alternate embodiments or locations may be indicated with use of prime (') superscripts, without further mention thereof. DETAILED DESCRIPTION OP THE INVENTION Referring now to the drawing, FIG. 1 depicts, in a vertical position, a radiation shield 10 fabricated according to my design, affixed in place by steel bands 12 about a section of pipe 14. As illustrated, a first layer S.sub.1 of shield portions is provided; these are identified as shield portions S.sub.1 (1), S.sub.1 (2), and S.sub.1 (3). A second layer S.sub.2 of shield portions is provided; these shield portions are identified as shield portions S.sub.2 (1), S.sub.2 (2), and S.sub.2 (3). Thus, it can be seen that the radiation shield 10 can be fabricated using a sequence of radiation shield portions that are provided in one or more layers. When a plurality of layers S.sub.1, through S.sub.N are provided, N is a positive integral number corresponding to the number of layers provided. In each of the layers S.sub.1 through S.sub.N, one or more shield portions may be provided. In each such layer S.sub.N, shield portions may be described by a sequence of shield portions S.sub.N (1) through S.sub.N (X), where X is a positive integer representing the number of shield portions in that layer. As is intuitively obvious in view of the specific example set forth, and by use of the various figures of the drawing, the location of any one radiation shield layer S.sub.N may generally be described relative to other shield layers therebelow, such as S.sub.N-1 for the shielding layer immediately below layer S.sub.N, or relative to other layers thereabove, such as layer S.sub.N+1 for the layer immediately above layer S.sub.N. I prefer to fabricate each radiation shield 10 before attachment to pipe 14. As can be seen in FIG. 1A, when my radiation shields are used on a pipe 14, shield portions S.sub.1 (1), etc., must be fabricated so that the inner diameter D.sub.1 I substantially conforms to the outer diameter D.sub.P O of pipe 14. Then, the inner diameter D.sub.2 I of shield portions (e.g., S.sub.2 (1)) which are used in the second layer S.sub.2 must substantially conform to the outer diameter D.sub.1 O of the shield portions (e.g., S.sub.1 (1)) which are used in the first layer S.sub.1 of the shield portions. As is evident from FIG. 1A, each of the radiation shield portions is provided in the shape of a segment of an annulus. For convenience in fabrication, I have prepared a table which eliminates the need to calculate diameter dimensional data for commonly encountered pipe sizes. In Table I below, I have provided the size and weight encountered for a first layer S.sub.1, shield portion and for a second layer S.sub.2 shield portion, when the radiation shield making up each layer is fabricated from a sheet TABLE 1 ______________________________________ LEAD CUT SIZES FOR FABRICATION OF HALF-ROUND SHIELDS Pipe Width - W Width - W Pipe Outside for 1/4" for 1/2" Weight Weight size Diameter (inner piece) (outer piece-) 1/4" .times. 12" 1/2" .times. 12 (in.) D.sub.p O S.sub.1 layer) S.sub.2 layer) pounds pounds ______________________________________ 1 1.365 21/2 27/8 31/8 63/4 2 2.375 41/8 5 51/8+ 113/8+ 3 3.5 57/8 63/4 73/8 151/2 4 4.5 71/2 81/4 93/8 193/4 5 5.563 91/8 97/8 113/8 233/4 6 6.625 103/4 12 133/8 277/8 8 8.625 131/2 15 167/8 351/2 10 10.75 171/4 181/2 211/2+ 45 12 12.75 203/4 22 26 531/2 ______________________________________ NOTE: This table assumes 1/4 (0.25) inch lead sheet stock is used. of lead S of thickness T, and where the thickness T is selected at 1/4 (0.25) inches. As indicated in FIG. 1B, my preferred method of fabricating a half-round coated lead shield, such as any of the shield portions shown in FIG. 1, is to first determine the pipe size of pipe 14, and then cut a flat lead sheet S of 0.25 inches thickness T into the width W indicated according to Table 1, and in a desired length L. Then, using a pipe mold 14', the flat lead sheet is molded to fit the pipe 14 size. When a final radiation shield 10 is to be made in two 0.25 inch layers, the inner layer S.sub.1 is first made, and then the second layer S.sub.2 is preferably shaped over the piece for the first layer S.sub.1. For ease in fabrication, I prefer to leave an offset gap G of about one (1) inch, to offset, layerwise, the gap between adjacent shield portions (e.g., S.sub.1 (1) to S.sub.1 (2) and S.sub.2 (1) and S.sub.2 (2). Space 20 between adjacent shield portions should be minimized in order to avoid loss of shielding effectiveness. To assemble shield portions into a final radiation shield 10, for example when fabricating a final radiation shield 10 of 1/2 (0.5) inch thickness (using one quarter inch lead in each of layers S.sub.1 and S.sub.2, or TS.sub.1 +TS.sub.2 !=1/2 inch) I prefer to use 1" deck screws 22, as can be seen in FIG. 3. The shield portions (e.g., S.sub.1 (1) and S.sub.2 (1)) are fastened together by running the deck screws 22 from outside to the inside. Preferably, pre-drilled holes are avoided. Use of deck screws 22 is important since they do not require pre-drilled holes, and the threads do not load with lead and tear out as they are run into the lead. Also, the head 24 of deck screws 24 are of a counter sunk type and they will run in flush with the outside or upper surface 26 of the outer layer lead shield portion being assembled. Also, desirable deck screws are provided in hard, brittle materials, which make it easy to break off the threads that protrude through the lead, for example when using one (1) inch screws with a one-half (1/2) inch total shield thickness. Preferably, the protruding part can be broken off with a hammer or snipped off with pliers, and the lead around the stump 28 can be shaped to assure there are no protruding sharp edges. Likewise, corners C of the shaped lead sheet S are rounded, usually with a hammer, to assure that there are no sharp corners on the finished radiation shield 10. After the shield portions of the radiation shield 10 are joined, the shield 10 is coated to provide a final cured coating 30. A preferred coating material used to cover the lead and provide a coating 30 is a flexibilized Bisphenol A epoxy which is cross linked with a modified cycloaliphatic amine curing agent. Ideally, such a coatings is provided as a two component, medium viscosity (1250 cps at 77.degree. F.) epoxy, with 100 percent solids. I prefer a product with minimal color fade or degradation upon exposure to sunlight, and with the following performance properties: (1) Tensile strength, using method ASTM D-538: 1100 pounds per square inch PA1 (2) Percent elongation, using method ASTM D-638 60 percent (minimum) PA1 (3) Shore D Hardness, using method ASTM D-2230 37 hardness PA1 (4) Tensile Shear Strength, method ASTM D-1002 347 pounds per square inch. The coating can be applied by roller, paint-brush, or by spray. Particular attention must be paid to the corners C and edges E, to avoid any thin or shallow coated areas. If touch up coats are used or required, they should meet the same coating specifications as the original coating. Once completed, the radiation shield 10 can be placed directly on piping, such as pipe 14, or can also be used as shadow shielding, particularly if it is provided in flat sheet form (as shown as shield 40 in FIG. 4 below) rather than in the shaped shield form just described above. The installation method chosen will also depend upon whether the installation is to be temporary or permanent. For ease of installation, I prefer to use grommets 36 which have in inside wall 37 to define a through passageway 38 in the shield 40. When grommets 36 are used, it is preferable to coat the lead sheet S with coating 30 first, and then install grommet 36. I prefer to grommet shields 40 with grommets 36 on twelve (12) inch centers, starting about six (6) inches from an outer edge E and centered about one and one-half (1 1/2) inch from the top Z. Grommets are installed using a five-eighth (5/8) inch hole punch and ideally, five-eighths (5/8) inch brass grommets 36 are utilized. This size allows for some slack when the shield 40 is placed using one-half (1/2) inch bolts. After installing grommets, if the coating 30 has been damaged, it should be repaired, prior to using the shield 40. If sheets 40 of coated lead radiation shield are to be provided as temporary shadow shielding, then the sheets 40 can be supported by scaffold tube framing 42, with the coated lead shield sheets 40 hung on S-hooks 44, as indicated at the bottom of FIG. 8 below. As noted in FIG. 8A, I prefer to make S-hooks 44 from about a 5/16 inch round stock, with about a 2 1/8 inch diameter in each arm of the S-hook 44, and an extension arm 45 to each end of the S-hook 44 of about three quarters (3/4) of an inch. If the coated lead sheet 40 is used for temporary pipe shielding, then half-rounds can be supported from pipe 14 by wire ties, instead of bands 12 shown above in FIG. 1. However, if the installation is for permanent shadow shielding, then support will be analogous to that shown for a stainless steel encapsulated shield 50 as shown in FIG. 12 below. Specifically, a structural steel support 52 is used to hold an attachment structural steel member 54, to which the shield (such as 40 instead of 50) can be affixed via fastener such as bolt 56 and nut (not shown). Alternately, a shield 40 can be permanently affixed on scaffold tube support frame with wire ties approved for use in the service environment. I prefer to use a a lead sheet S sized 23.5 inches by 47.5 inches, for the normally encountered radiation shielding situations. Such size sheets S are also advantageous for manufacture of full size twenty four (24) inch by forty eight (48) inch radiation shields 50 which are encapsulated with stainless steel 60. By using the suggested lead sheet S size, space is allowed for covering the lead sheet S and riveting the stainless steel 60, so that the completed panel dimensions are not greater than twenty four (24) inches by forty eight (48) inches. Turning now to FIGS. 5, 6, and 7, one convenient method for manufacture of my stainless steel encapsulated lead shields is shown. Typically, I find that a 20 gauge stainless steel sheet 60 is adequate to provide the encapsulation that my radiation shields 50 require. First, an inner layer of at least one sheet S of solid radiation shielding material, preferably lead, is provided, cut to desired size as described herein. Then, a first, obverse stainless steel panel 62 is cut to a desired size. In FIG. 5, taken looking down at the left side flange 64 on the obverse side 65, shows how flange 64 preferably extends from the face 67 of obverse side 65 at at about a ninety (90) degree angle therefrom. A companion left side flange 66 extends also at preferably about a ninety (90) degree angle from the face 68 of the reverse side 69. This FIG. 5, in combination with FIG. 6, shows the method which is used for encasing the the right and left sides of a radiation shield 50 which has a stainless steel outer casing 60. Bottom 70 of the radiation shielding material S is at the left of FIG. 5. Left edge 72 the radiation shielding material S is at the top of FIG. 5. Preferably matching size apertures A defined by edge portions 80 are located in flanges 64 and 66 are provided for use in securing fasteners R thereto. Note that flanges 64 and 66 extend above edge 72 of the radiation shielding material by a distance I, to leave a void of width I, which void is filled with a suitable sealant 74. I prefer to use silicon caulking or adhesive, but in some applications, a polyvinylchloride type filler will also be acceptable. Preferably after filling the void with sealer 74, flanges 64 and 66 are joined with a mechanical fastener, typically pop type rivets R. I like to use rivets R on one inch centers along the edge of the radiation shield. In any event, sufficient space must be provided, i.e., width I, for the rivets R or other fastening device to be finally assembled without intruding into the at least one sheet S of solid radiation shielding material. Similar construction is typically used for both sides, and generally is also desirable along the bottom of shield, where an obverse bottom flange and a reverse bottom flange are joined in similar fashion. At the top, in order to provide an extra measure of protection against intrusion of water or steam, a U-shaped cap 80 is preferably provided. Just as with construction of the sides, a void of height K is provided to allow fasteners to join parts without intruding into the at least one sheet S of solid radiation shielding material. The obverse side 65 has a top flange 82, and the reverse side 69 has a top flange 84. Cap 80 in the shape of an elongate, a generally U-shaped channel having a reverse side leg 86 and an obverse side leg 88. The cap 80 is fitted downward over top flange 84 of the reverse side 69, and downward over the top flange 82 of the obverse side 65. The cap 80 is fitted in a manner where an inner portion 90 of the reverse side leg 86 is placed in an abutting relationship with the face 68 of the reverse side 69. Likewise, the cap 80 is fitted in a manner where an inner portion 92 of the obverse side leg 88 is placed in an abutting relationship with the face 67 of the obverse side 65. In this way, a first mechanical fastening device R.sub.1 is used to join the reverse side leg 86 of the cap 80 to said reverse side 69. A second mechanical fastening device R.sub.2 is used to join the obverse side leg 88 of the cap to the obverse side 65. It is important that the mechanical fastening devices R.sub.1 and R.sub.2 do not intrude into the at least one sheet S of radiation shielding material. Ideally, the flanges are formed in an integral, one-piece fashion with each of the obverse and reverse panels used to encapsulate the lead sheet used as a core. As further illustrated in FIG. 7, I prefer to provide a grommet 36' in my radiation shield 50. A plurality of grommets 36' defines through passageways in the radiation shield, so that the radiation shield 50 may be upheld by a supporting structure protruding through said grommet. An ideal supporting structure is provided by my hangers 100 and 110, as illustrated in FIGS. 9 and 10, respectively. As shown in FIG. 9, hanger 100 includes an elongate, flat bar portion 102, a backwardly curved hook portion 104, and a forwardly protruding stud 106. The stud 106 is attached to the elongate flat bar portion 102 at the upper reaches 108 thereof, and extends forwardly therefrom in a generally horizontal manner, so that the protruding stud 106 extends forward a distance to support, hanging vertically therefrom, a radiation shield 50. As shown in FIG. 10, an alternate hanger 110 is set forth. Hanger 110 is similar to the hanger 100 just described, but further includes a J-shaped hook 112 at the lower reaches 114 thereof. The J-shaped hook 112 is preferably formed as an integral part of the elongate flat bar portion 102, and is provided in sufficient width and shape to cradle therein a planar lower edge 116 of a shield 50 which may require support of stabilization. FIG. 10A shows a side view of the J-hook hanger. Radiation shields using my design can be custom manufactured to be installed around pipe, conduit, or other structures from which radiation is being emitted. The exact design of the shielding will be based on the radiation source(s), the dose rate both (i) contact and (ii) general area type, the project shielding requirements (whether job specific or area dose rate reduction driven), the area configuration, including environmental conditions, the duration (temporary or permanent), and various engineering requirements, such as structure loading and seismic requirements. In any event, it will thus be seen that the objects set forth above, including those made apparent from the proceeding description, are efficiently attained, and, since certain changes may be made in carrying out the construction of a radiation shielding apparatus to generally in the manner described, and while still achieving the objectives as set forth herein, it is to be understood that the invention may be embodied in other specific forms without departing from the spirit or essential characteristics thereof. For example, while I have set forth exemplary designs for a stainless steel encapsulated radiation shield design, many other embodiments are also feasible to attain the result of the principles of the apparatus and via use of the methods disclosed herein. Therefore, it will be understood that the foregoing description of representative embodiments of the invention have been presented only for purposes of illustration and for providing an understanding of the invention, and it is not intended to be exhaustive or restrictive, or to limit the invention to the precise forms disclosed. On the contrary, the intention is to cover all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as expressed in the appended claims. As such, the claims are intended to cover the structures and methods described therein, and not only the equivalents or structural equivalents thereof, but also equivalent structures or methods. Thus, the scope of the invention, as indicated by the appended claims, is intended to include variations from the embodiments provided which are nevertheless described by the broad meaning and range properly afforded to the language of the claims, or to the equivalents thereof.
048790877	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring now to FIG. 1, a description will be given of a BWR power plant in accordance with a preferred embodiment of the present invention. A reactor core 2 is disposed in a tubular core shroud 3 provided in a reactor pressure vessel 1, and is loaded with a multiplicity of fuel assembles. A plurality of control rods 4 are inserted into the core 3, and a plurality of control rod driving apparatuses 5 are connected to the respective control rods 4. A plurality of jet pumps 6 having throat portions 7 are disposed in an annular space formed between the reactor pressure vessel 1 and the core shroud 3. The throat portions 7 are located above the core 2. A feed water sparger 8 is disposed above the throat portions 7 in the reactor pressure vessel 1 and has a header and a multiplicity of injection nozzles for injecting feed water. The injection nozzles are disposed in the header and are oriented toward the jet pumps. The feed water sparger 8 is located below the position at which a feed water pipe 33 is installed in the reactor pressure vessel 1. This is to ensure that low-temperature feed water injected from the feed water sparger 8 can be sucked by the jet pumps 6 at as practically low a temperature as possible. A tubular riser portion 9 is installed on the upper end of the core shroud 3. The diameter of the riser portion 9 is the same as that of the core shroud 3. The upper end of the riser portion 9 is located above the installation level of the feed water sparger 8. The feed water sparger 8 is disposed around the riser portion 9. A steam separator 10 is installed above the riser portion 9 via a stand pipe 11. A description will now be given of the construction of turbines and their peripheral parts. The shafts of a high-pressure turbine 13, a low-pressure turbine 14, and a generator 15 are connected to each other. However, the generator 15 is detachably connected to the low-pressure turbine 14. The high-pressure turbine 13 is connected to the reactor pressure vessel 1 by means of a main steam pipe 17, while the low-pressure turbine 14 is connected to the high-pressure turbine 13 by means of a main steam pipe 18. A condenser 16 is connected to a steam discharge port (not shown) of the low-pressure turbine 14. A steam stop valve 19 and a steam regulator valve 20 are provided in the main steam pipe 17. One end of a bypass pipe 21 having a bypass valve 22 is connected to the main steam pipe 17 upstream of the steam stop valve 19, while the other end thereof is connected to the condenser 16. A feed water pipe 30 is connected to the condenser 16. A condenser pump 23, a condensate demineralizer 24, a plurality of low-pressure feed water heaters 25, and a plurality of high-pressure feed water heaters 26 are disposed in this feed water pipe 30 in that order from the condenser 16 side toward the reactor pressure container 1 side. The feed water pipe 30 is branched off at a branching point 34 downstream of the high-pressure feed water heaters 26 into a feed water pipe 31 (pipe for jet pump driving water) and the feed water pipe 33 (supply pipe for the feed water sparger). A feed water pump 28, a final-stage high-pressure feed water heater 27, and a shut-off valve 35 are provided consecutively in the feed water pipe 31. In FIG. 1, the feed water pump 28 which is disposed downstream of the branching point 34 and downstream of the high-pressure feed water heater 27 may be disposed upstream of the high-pressure feed water heater 27, as shown in FIG. 2. If the feed water pump 28 is disposed upstream of the high-pressure feed water heater 28, as shown in FIG. 2, the pressure of feed water applied to the suction side of the feed water pump 28 becomes equal with that of feed water applied to the suction side of a feed water pump 29 which will be described later. This is because a pressure loss due to the high-pressure feed water heater 27 is capable of avoiding the effect exerted on the suction side of the feed water pump 28. The feed water pipe 31 is led to the reactor pressure vessel 1, and a tip thereof is constituted by a nozzle 32 which is open above the throat portion 7. The feed water sparger 8 is located above the nozzle 32. The other feed water pipe 33 is connected to the feed water sparger 8. The feed water pump 29 is disposed in the feed water pipe 33. The low-pressure feed water heater 25 and the high-pressure heat water heater 26, 27 respectively introduced part of the steam generated in the reactor pressure vessel 1 so as to heat the feed water. In other words, a space inside the shell of the low-pressure feed water heater 25 is communicated with the low-pressure turbine 14 by means of a bleeding pipe 36 having a bleeding valve 37. A space inside the shell of the high-pressure feed water heater 27 is communicated with the high-pressure turbine 13 by means of a bleeding pipe 40. A bleeding valve 39 is provided in the bleeding pipe 38, and a bleeding valve 41 in the bleeding pipe 40. Each of the spaces inside the shells of the low-pressure feed water heater 25 and the high-pressure feed water heaters 26, 27 is communicated with the condenser 16 by means of a drain pipe 42. Further, the feed water pipes 31, 33 are communicated with each other by means of a bypass pipe 43 which is provided with a shut-off valve 44. One end of the bypass pipe 43 is connected to the feed water pipe 33 downstream of the feed water pump 29, while the other end thereof is connected to the feed water pipe 31 between the feed water pump 28 and the shut-off valve 35. Incidentally, the following are installed as detectors: a level gauge 50 for detecting the reactor water level L.sub.R in the reactor pressure vessel 1; a flow meter 51 for detecting a flow rate of feed water (a flow rate of jet pump driving water, W.sub.J), which constitutes driving water for the jet pump 6; a flow meter 52 for detecting a flow rate of feed water (a flow rate of the feed water sparger, W.sub.S) in the feed water pipe 33 which is supplied from the feed water sparger 8 to the reactor pressure vessel 1; and a flow meter 53 for detecting a flow rate of main steam (W.sub.M) flowing through the main steam pipe 17. The level gauge 50 is installed in the reactor pressure vessel 1, the flow meter 51 in the feed water pipe 31, the flow meter 52 in the feed water pipe 33, and the flow meter 53 in the main steam pipe 17. Further installed are a thermometer 54 installed in the feed water pipe 31 and adapted to detect a temperature (T.sub.J) of feed water (jet pump driving water); a thermometer 55 installed in the feed water pipe 33 and adapted to detect a temperature (T.sub.S) of feed water (water in the feed water sparger); an output detector 56 installed in the core 2 and adapted to detect a reactor output (Q.sub.R); a pressure gauge 57 installed in the main steam pipe 17 and for detecting the main steam pressure (PS); and a tachometer 58 for detecting the number of revolutions of the turbine (N.sub.G). Each of the amounts of status detected by the level gauge 50, flow meters 51, 52, and 53, thermometers 54, 55, and reactor power detector 56 are transmitted to the water level/reactor power controller 70, which controls the feed water pumps 28, 29 and shut-off valves 35, 44. Each of the amounts of status detected by the pressure gauge 57 and the tachometer 58 is transmitted to the turbine controller 80, which controls the steam stop valve 19, steam regulator valve 20, and bypass valve 22. Installed as other controllers are the feed water temperature controller 46 for controlling an amount of opening of the bleeding valve 41 on the basis of output signals of the temperature gauges 54, 55, a bleeding valve controller 90 for controlling the amounts of opening of the bleeding valves 37, 39 on the basis of output signals of the turbine controller 80, and a control rod driving apparatus controller 100 for controlling the control rod driving apparatuses 5. Of signals output from the feed water temperature controller 46 and the bleeding valve controller 90, a signal selector 47 preferentially transmits the former output signal to the bleeding valve 41. Furthermore, the general controller 60 effects the general control of the water level/reactor power controller 70, turbine controller 80, bleeding valve controller 90, and control rod driving apparatus controller 100. The operation of the BWR power plant having the above-described arrangement in accordance with this embodiment will be described below. The control of reactor power in this embodiment is effected by adjusting the insertion of the control rods 4 into the core and removal thereof as well as a core flow rate (a flow rate of cooling water fed to the core 2). Steam generated in the core 2 rises in the riser portions 9 and is introduced into the steam separator 10, which separates cooling water contained in the steam. The separated cooling water flows downwardly toward an annular descending flow passage 45 formed between the riser portions 9 and the reactor pressure vessel 1. Meanwhile, the steam in which cooling water is removed by the steam separator 10 is introduced into the high-pressure turbine 13 through the main steam pipe 17 and further into the low-pressure turbine 14 through the main steam pipe 18. During the normal operation of the plant, the steam stop valve 19 and the steam regulator valve 20 are open, while the bypass valve 22 is closed. The steam rotates the high and low-pressure turbines 13, 14. Also, the generator 15 which is connected to these turbines is rotated simultaneously. The steam exhausted from the low-pressure turbine 14 into the condenser 16 is condensed into water in the condenser 16. This condensed water is supplied again to the reactor pressure vessel 1 as feed water through the feed water pipes 30, 31, and 33. Namely, the pressure of the condensed water, i.e., feed water, in the condenser 16 is increased, and after it is purified by the condensate demineralizer 24, the feed water reaches the branching point 34 through respective heat conductive pipes of the low-pressure feed water heaters 25, 26. The steam respectively extracted from the low- and high-pressure turbines 14, 13 by the bleeding pipes 36, 38 is supplied to spaces inside the shells of the low- and high-pressure feed water heaters 25, 26. The flow rates of the extracted steam are controlled by the bleeding waves 37, 39. The feed water introduced into the heat conductive pipes of the feed water heaters 25, 26 is heated by the extracted steam supplied to the spaces inside the shells, and its temperature is raised as a result. The extracted steam is condensed in the spaces inside the shells upper heating of the feed water, and is thereby converted into a drain. This drain is returned to the condenser 16 by a drain pipe 42. Some portion of the feed water which has reached the branching point 34 passes through the feed water pipe 33 and its pressure is increased by the feed water pump 29. This feed water is then discharged from the feed water sparger 8 into the annular descending flow passage 45 in the reactor pressure vessel 1. The cooling water discharged from the feed water sparger 8 also contributes to the adjustment of the water level inside the reactor pressure vessel 1, as will be described later. On the other hand, the pressure of the remaining portion of the feed water which has reached the branching point 34 is increased by the feed water pump 28, and its temperature is increased after it is introduced into the heat conductive pipe inside the high-pressure feed water heater 27. This feed water is then discharged from the nozzles 32 into the jet pump 6. As this feed water is discharged from the nozzle 32 into the jet pumps 6, the cooling water remaining in the annular descending flow channel 45 is sucked into the jet pumps 6. The high-pressure steam extracted from the high-pressure turbine 13 by the bleeding pipe 40 is supplied to the space inside the shell of the high-pressure feed water heater 27. The bleeding pipe 40 extracts steam from the high-pressure turbine 13 at an upstream position as compared to that at which the bleeding pipe 38 does so. Accordingly, the temperature of the steam extracted by the bleeding pipe 40 is higher than that of the steam extracted by the bleeding pipe 38. The extracted steam introduced into the space inside the shell of the high-pressure feed water heater 27 heats the feed water flowing in the conductive pipe and is condensed into a drain. This drain is also introduced into the condenser 16 through the drain pipe 42. The feed water flowing through the feed water pipe 31 contributes as the driving water for the jet pump 6. The feed water discharged from the nozzle 32 as well as the cooling water in the annular descending flow passage 45 which is sucked into the jet pumps 6 are discharged from the jet pumps 6 in a mixed state. The cooling water discharged from the jet pumps 6 is introduced into the core 2 from therebelow. The embodiment described above effects (I) the use of two different kinds of feed water having different temperatures, (II) the control of the core flow rate (the control of outputs with finer adjustment than that of the control rods) and the control of the water level by means of the feed water, (III) the control of protection at the time of an emergency, and (IV) the control of response characteristics of generator outputs. A detailed description will be given hereafter of these items in sequence. (I) Use of Two Different Kinds of Feed Water Having Different Temperatures In this embodiment, a portion of the feed water branched upstream of the high-pressure feed water heater 27 is supplied from the feed water sparger 8 into the reactor pressure vessel 1 without passing through the high-pressure feed water heater 27, while the remaining portion of the feed water is supplied into the reactor pressure vessel 1 (specifically into the jet pumps 6) through the high-pressure feed water heater 27. Accordingly, the temperature of the feed water sparger 8 is lower than that of the feed water supplied from the feed water pipe 31 into the jet pumps 6 as the driving water by a portion which is not heated by the high-pressure feed water heater 27. To put is conversely, the temperature of the feed water which is used as the driving water for the jet pump 6 is higher than that of the feed water supplied to the feed water sparger 8 by a portion which is heated by the high-pressure feed water heater 27. The temperature difference between these two kinds of feed water is adjusted to a predetermined value through the control of the amount of opening of the bleeding valve 41 by the feed water temperature controller 46. As shown in FIG. 3, the feed water temperature controller 46 has a temperature estimator 46A, adders 46B, 46C, a controller 46D, and a function generator 46E. The temperature estimator 46A is used to determined a temperature T.sub.JSC of a suction current 45A (FIG. 45A) sucked into the jet pumps 6 on the basis of the temperature T.sub.S of the feed water inside the feed water pipe 33 (hereafter referred to as the feed water sparger water) measured by the thermometer 55. A difference .DELTA.T between the temperature T.sub.JSC and the temperature T.sub.J of the feed water inside the feed water pipe 31 (hereafter referred to as the jet pump driving water) measured by the thermometer 54 is determined by the adder 46B. The function generator 46E stores the characteristics L.sub.3 of a relation between a reactor power Q.sub.R and a set temperature difference .DELTA.T.sub.0. The set temperature difference .DELTA.T.sub.0 is determined in correspondence with the reactor power Q.sub.R. The characteristics L.sub.3 are located below an alarm issuing line L.sub.2 shown in FIG. 10, which will be described below, and are set in such a manner that the temperature difference between the jet pump driving water and the feed water sparger water assumes a necessary value. The function generator 46E outputs the set temperature difference .DELTA.T.sub.0 corresponding to the reactor power Q.sub.R measured by the reactor power detector 56. The set temperature difference .DELTA.T.sub.0 output from the function generator 46E as well as the temperature difference .DELTA.T output from the adder 46B are input to the adder 46C. The adder 46C determines a deviation T.sub.X between the set temperature difference .DELTA.T.sub.0 and the temperature difference .DELTA.T. The controller 46D outputs an opening-amount signal ST.sub.1 for the bleeding valve 41 on the basis of the deviation T.sub.X. The opening-amount signal ST.sub.1 is transmitted to the signal selector 47, which transmits the opening-amount signal ST.sub.1 in preference to an opening-amount signal ST.sub.2, i.e., an output of the bleeding valve controller 90. When the opening-amount signal ST.sub.2 is not output or delivered, i.e., when the opening-amount signal ST.sub.2 is a signal which corresponds to a signal V.sub.2 whose level is 0, as will be described below, the opening-amount signal ST.sub.1 is selected and transmitted to the bleeding valve 41. In short, in cases other than when a bleeding valve opening controller 91 (FIG. 14) of the bleeding valve controller 90 operates (i.e., when a deviation signal U, which is an output of an adder 91A shown in FIG. 14, is negative), the amount of opening of the bleeding valve 41 is adjusted by the opening-amount signal ST.sub.1 which is the output of the feed water temperature controller 46. Thus, since the feed water sparger water having a temperature lower than that of the jet pump driving water discharged from the nozzle 32 by a predetermined temperature difference is supplied from the feed water sparger 8 into the reactor pressure vessel 1, the temperature of the cooling water descending through the annular descending flow passage 45 declines, and, hence, low-temperature cooling water is sucked into the jet pumps 6 by means of the feed water injected from the nozzle 32. As a result, it is possible to prevent the occurrence of cavitation in the jet pumps 6. The aforementioned temperature difference between the two kinds of feed water is controlled in such as manner as to satisfy a predetermined value during normal operation. However, it does not mean that the temperature of the feed water flowing through the feed water pipe 31 can be increased noticeably insofar as the temperature difference between the two kinds of feed water satisfies the predetermined value. An important matter in this respect is that the temperature of the feed water inside the feed water pipe 31 is determined on the basis of the enthalpy of the cooling water at a core inlet which is required in the light of the core characteristics and in consideration of the enthalpy of the cooling water separated by the steam separator 10 and flowing downwardly through the annular descending flow passage 45 and the enthalpy of the feed water in the feed water pipes 31, 33. The thus-determined feed water temperature in the feed water pipe 31 is adjusted when the amount of opening of the bleeding valve 41 is controlled in such a manner that the temperature of this feed water becomes higher than that of the feed water in the feed water pipe 33 by a predetermined value. As has been described above, in accordance with this embodiment, since the feed water flowing through the feed water pipes 31, 33 is provided with a temperature difference (the temperature of the feed water in the feed water pipe 33 being lower), it is unnecessary to provide a heat exchanger inside the reactor pressure vessel 1 as has been required in the case of a conventional example, so that the internal structure of the reactor pressure vessel 1 can be simplified remarkably. Accordingly, the work concerning installation of in-reactor structures in the reactor pressure vessel 1 can be facilitated appreciably. Furthermore, it is possible to dispense with maintenance work concerning the heat exchanger inside the reactor pressure vessel 1 radioactivated to a high level, which has been necessary in the conventional example. In addition, since, in this embodiment, the high-pressure feed water heater 27, which is disposed outside the reactor pressure vessel 1 and is necessary for heating the feed water in the conventional example as well, is heated or unheated, it is possible to enlarge the temperature difference between the feed water in the feed water pipe 31 and that in the feed water pipe 33. Hence, it is possible to remarkably lower than temperature of the cooling water sucked into the jet pumps (i.e., the cooling water inside the annular descending flow passage 45) than that of the conventional example. Accordingly, in this embodiment, the extent to which the occurrence of cavitation in the jet pumps 6 can be prevented increases substantially, so that it is possible to substantially alter the reactor power by the control of the core flow rate. In this embodiment, therefore, it is readily possible to implement the load following operation, so that it is possible to effect the load following operation to a remarkable degree by the control of the core flow rate. Incidentally, the temperature difference between the jet pump driving water and the cooling water in the annular descending flow passage is very small in the conventional example. Consequently, the cooling effect due to the cooling water in the annular descending flow passage obtained by the jet pump driving water using the heat exchanger installed in the reactor pressure vessel is small, and a temperature drop of the cooling water sucked into the jet pumps is also small. In the conventional example, therefore, the extent to which the occurrence of the jet pump cavitation can be prevented is narrower than in this embodiment, so that it is impossible to cope with the occurrence of cavitation during a substantial change in the reactor power. Furthermore, since the feed water pipe 33 is branched upstream of the high-pressure feed water heater 27, which is an important apparatus for the feed water system, a cooling apparatus is not required for setting the temperature of the feed water inside the feed water pipe 33 lower than that of the feed water inside the feed water pipe 31. In addition, it is unnecessary to install a new heater for heating the feed water in the feed water pipe 31. In this embodiment, it is unnecessary to install a recirculation system with a recirculation system pipe conventionally disposed in a reactor container surrounding the reactor pressure vessel 1 or an internal pump having a complicated structure. In addition, the feed pumps 28, 29 are disposed outside the reactor container. Therefore, the reactor container can be made compact, and the structure of the interior of the reactor container can be simplified. Since the recirculation system and the internal pump become unnecessary and the maintenance thereof becomes unnecessary, a period of maintenance of the nuclear plant can be shortened. Although, in this embodiment, the feed water pipe is branched upstream of the final-stage high-pressure feed water heater 27, with respect to a BWR power plant having a different value of the rated output, the feed water pipe 33 may be branched off from the feed water pipe 30 upstream of the feed water heater two or three stages upstream of the final-stage high-pressure feed water heater 27. As a result, it is possible to remarkably lower the temperature of the feed water in the feed water pipe 33 below that of the feed water in the feed water pipe 31 as compared with the above-described embodiment. The usage of two different kinds of feed water having different temperatures in this embodiment has been conceived as a result of a study which is described below. In accordance with the conventional core nuclear design, a core flow rate W.sub.C at which the cooling water is supplied through the recirculation system is equivalent to about seven times as large as a total feed-water flow rate W.sub.FW (which is equivalent to a main steam flow rate W.sub.MS during normal operation). Accordingly, if the conventional core design is used as reference, even if the total feed water is alloted to the jet pumps 6 as the driving water, the jet pumps would require an M ratio of about 6.0, the M ratio being determined by the following formula: ##EQU1## Actually, however, it is necessary to divert a portion of the flow rate of the total feed water to the feed water sparger 8 at the branching point 34 so as to control the water level of the reactor. This is because it is necessary to control the water level of the reactor independently of other variables. Let us assume that one third of the total quantity of feed water is alloted to the feed water sparger 8 and the remaining two thirds to the jet pumps 6. If the flow rate of the feed water sparger water is W.sub.S and the flow rate of the jet pump driving water is W.sub.J, and if the core nuclear design is the same, the following formula holds: ##EQU2## Accordingly, an M ratio required becomes 9.5. Incidentally, the M ratio of conventional BWR jet pumps having a recirculation system is 2.5 or thereabout at the time of a rated output. Basically, the following four measures are conceivable to increase the M ratio: (1) The pressure of the driving water for the jet pumps is increased, and the flow rate of the driving water flowing through the throat portions of the jet pumps is also increased. PA1 (2) The required delivery pressure of the jet pumps is made small. PA1 (3) The temperature of a suction flow sucked by the jet pumps is decreased. PA1 (4) Multi-stage jet pumps are used. The characteristics of a jet pump will now be shown with reference to FIG. 4. The suction flow 45A (the flow of cooling water through the annular descending flow passage 45) is sucked into the throat portion 7 by virtue of the flow 31A of the jet pump driving water, and after the flows are mixed, the mixture is introduced into an outlet 6B of the jet pump 6 via a diffuser 6A of the jet pump 6. FIG. 4 further shows changes in the pressure P.sub.S of the suction flow 45A and a pressure P.sub.D of the flow of the jet pump driving water that take place in the direction of the flow. The pressure of the two flows assumes a minimum pressure P.sub.M at the throat portion, and the pressure of the flow subsequently increases due to the effect of the diffuser 6A, and a delivery pressure P.sub.O is obtained at the outlet 6B. If the above-described measure (1) is adopted and the driving water flow 31A with an appreciably large pressure P.sub.D, namely, with an appreciably high flow rate is used, the pressure of the suction flow 45A becomes lower than a saturation pressure at the throat portion 7 (this phenomenon being called cavitation). When cavitation takes place, the efficiency of the jet pump 6 is remarkably deteriorated, possibly causing damage to the equipment. Accordingly, it is essential to enlarge the M ratio while ensuring that the cavitation will not occur. One method of increasing the M ratio without the occurrence of the cavitation in the jet pump is to sufficiently lower the temperature of the suction flow 45 and to cause the same not to boil at the pressure P.sub.M since the occurrence of the cavitation is attributable to the boiling which takes place due to the fact that the pressure P.sub.M of the throat portion 7 is lower than the saturation pressure of the suction flow 45A. In other words, this involves the adoption of the above-described method (3). A conventional example based on this idea is disclosed in the aforementioned Japanese Patent Examined Publication No. 43-23117 entitled "Reactor System". As a means of lowering the temperature of the suction flow in this conventional example, a group of heat conductive pipes of the heat exchanger are connected to the feed water pipe 31, shown in FIG. 4, in the reactor pressure vessel 1, and heat exchange is effected between the suction flow and the jet pump driving water which is supplied by this feed water pipe. In this conventional example, the temperature difference between the suction flow and the jet pump driving water is very small, and since a sufficient length and size cannot be provided for the heat exchanger, there has been problems in practical use. Incidentally, this conventional example employs the above-described measure (4) as well. Accordingly, upon examining various practical measures, the present inventors reached a conclusion that it suffices if the temperature of the feed water supplied from the fed water sparger 8 is lowered sufficiently below that of the feed water used as the jet pump driving water. It should be noted that, in this embodiment, the above-described measure (2) is also adopted wherein the delivery pressure P.sub.O of the jet pump 6 is made small. It is necessary to make the delivery pressure P.sub.O greater than a pressure loss (static pressure plus dynamic pressure) based primarily on the core 2, a core outlet chamber (plenum), the steam separator 10, and the stand pipe 11. For this reason, a chamber in which a core outlet chamber is elongated upwardly of the core 2 (a chamber which is elongated up to the vicinity of the feed water sparger 8), i.e., the riser portion 9, is provided. The riser is also called a chimney. As a result, a large quantity of the two-phase current exists above the core 2, and a natural circulating force based on the buoyancy of void increases. Since the cooling water is driven by this natural circulating force, it is possible to reduce the delivery necessary pressure for the jet pump 6 as compared with a case where the riser portion 9 is not provided. The delivery pressure shown in FIG. 4 is determined by the magnitude of the pressure P.sub.O of the flow 31A of the jet pump driving water. If the delivery pressure P.sub.O is sufficiently small, it is possible to obtain a large M ratio even if the jet pump 6 is not driven by the jet pump driving water having a very large pressure P.sub.O. (II) Control of the Core Flow Rate and Water Level Control by Feed Water The control of the core flow rate and the water level in this embodiment is effected by the water level/reactor power controller 70. As shown in FIG. 5, the water level/reactor power controller 70 comprises a water level controller 71, a reactor power master controller 72, and a plant status discriminator 73. The arrangement of the water level controller 71 is shown in FIG. 6. The water level controller 71 has an adjuster 71A, a switch 71B, and adders 71C, 71D, 71E, and 71F. Input to the water level controller 71 are the reactor water level L.sub.R measured by the level gauge 50, the flow rate W.sub.J of the jet pump driving water measured by the flow meter 51, the flow rate W.sub.S of the feed water sparger water measured by the flow meter 52, the main steam flow rate W.sub.M measured by the flow meter 53, and a three element/one element control changeover signal S.sub.9A output from the general controller 60. The switch 71B is closed when the three element/one element control changeover signal S.sub.9A instructs three element control, while the switch 71B is opened when the signal S.sub.9A instructs one element control. The three element/one element control changeover signal S.sub.9A instructs one element control until the turbines 13, 14 reach the reactor power (an approx. 10% power). The adder 71C determines a deviation between the input reactor water level L.sub.R and a set value L.sub.RO of the reactor water level. The adder 71E adds the flow rate W.sub.J of the jet pump driving water and the flow rate W.sub.S of the feed water sparger water. The adder 71F determines a deviation between an output signal of the adder 71E and the main steam flow rate W.sub.M. When the switch 71 is closed on the basis of the three element/one element control changeover signal S.sub.9A (instructing the control of three elements), a deviation signal output from the adder 71F is input to the adder 71D. When the switch 71B is open on the basis of the three element/one element control changeover signal S.sub.9A (instructing the control of one element), the deviation signal output from the adder 71F is no input to the adder 71D. The adder 71D inputs the signals output from the adders 71C, 71F when the switch 71B is closed (when the reactor power exceeds 10%), and adds the respective signals. When the switch 71B is open (i.e., when the reactor power is below 10%), the adder 71D outputs the output signal of the adder 71C as it is. The adjuster 71A outputs a feed-water sparger flow-rate request signal W.sub.SO in response to the output signal of the adder 71D such that the reactor water level becomes the set value L.sub.RO of the reactor water level. This feed-water sparger flow-rate request signal W.sub.SO becomes the output of the water level controller 71 and is input to the plant status discriminator 73. The reactor power master controller 72 has the arrangement shown in FIG. 7. In other words, the reactor power master controller 72 has a switch 72B which switches over the adjuster 72A to a fixed terminal A or B. Input to the reactor power master controller 72 are a load change request signal .epsilon..sub.LOAD, a manual load change request signal .epsilon..sub.M, and a manual/automatic changeover signal S.sub.9B which are output from the general controller 60. These load request signals represent a difference between an actual reactor power and a targeted reactor power (or a difference between an actual output of electricity of the generator 15 and a targeted output of electricity). The load change request signal .epsilon..sub.LOAD is input to the fixed terminal A. The manual load change request signal .epsilon..sub.M is input to the fixed terminal B. The switch 72B is connected to the fixed terminal B when the manual/automatic changeover signal S.sub.9B instructs manual. The load change request signal .epsilon..sub.LOAD or .epsilon..sub.M is input to the adjuster 72A, a jet pump driving water flow-rate request signal W.sub.JO is output in such a manner that both load change request signals become zero. This request signal W.sub.JO is input to the plant status discriminator 73. When the BWR power plant is operating normally, the plant status discriminator 73 shown in FIG. 8 outputs both values of the input feed water sparger flow-rate request signal W.sub.SO and the jet pump driving water flow-rate request signal W.sub.JO as they are as the feed water sparger flow-rate request signal W.sub.SP and the jet pump driving water flow-rate request signal W.sub.JD. The detailed arrangement and functions of the plant status discriminator 73 will be described in the ensuring item (III). The feed water sparger flow-rate request signal W.sub.SP is input to the feed water pump 29. The feed water pump 29 changes the number of revolutions in correspondence with the feed water sparger flow-rate request signal W.sub.SP and thereby adjusts the flow rate of the feed water flowing through the feed water pipe 33 such that the reactor water level L.sub.R becomes the set value L.sub.R of the reactor water level. As a result, it is possible to maintain the reactor water level L.sub.R at a predetermined level regardless of the reactor output. The jet pump driving water flow-rate request signal W.sub.J is input to the feed water pump 28. The feed water pump 28 changes the number of revolutions in correspondence with the jet pump driving water flow-rate request signal W.sub.JD, and adjusts the flow rate of the feed water flowing through the feed water pipe 31 such that the reactor power Q.sub.R becomes a predetermined reactor power. The control of the flow rate of the feed water in the feed water pipe 31 subjects the reactor flow of the cooling water discharged from the jet pump in the same way as a conventional recirculating system having a recirculating pipe, thereby leading to the control of the reactor power. The fine control of the reactor power based on a change in the core flow rate due to the feed water in the feed water pipe 31 in this embodiment is effected in a reactor power control area (a high reactor power area exceeding a reactor power for starting a Pellet) Clad Mechanical Interaction (PCMI) disclosed in the specification of Japanese Patent Examined Publication No. 57-11038. The output control in an area below the output for starting the PCMI is effected by the control rods 4 in the same way as described in said Japanese Patent Examined Publication No. 57-11038. A change in the reactor power through the control of flow rate of the driving water in the jet pump in accordance with this embodiment is identical as the one encountered in the conventional BWR power plant having a recirculating system. However, the conventional recirculating system employs the flow of jet pump driving water having substantially the same temperature as that of the suction flow in the jet pumps. In contrast, in this embodiment, the temperature of the core flow rate undergoes slight change upon being influenced by the effect of a temperature change (approx. 1/10 if the M ratio is 10) of the jet pump driving water flow 31A, so that this slight change further changes the core void reactivity. Accordingly, as shown in FIG. 9, the variation characteristics of the reactor power in this embodiment slightly differ from a case where a conventional recirculating system is used. The variation characteristics of the former assume a higher level than that of the latter. (III) Protective Control During Emergency In the BWR power plant having the conventional recirculating system, the control of the reactor water level is effected using the flow rate W.sub.FW of the total feed water. In this embodiment, however, a major portion of the feed water (2.W.sub.FM /3) is used for the control of the reactor output discharged into the jet pumps 6, while the remaining portion of the feed water (W.sub.FW /3) is used for the control of the reactor water level. Accordingly, the capability of adjusting the reactor water level is smaller than that of the BWR power plant having the conventional recirculating system. This does not present problems during normal operation. However, when the plant status of the BWR power plant involves a sudden and substantial change and when that change is abnormal, there is a possibility that a substantial decrease occurs in the reactor water level. It is necessary to avoid the occurrence of such a phenomenon. For this reason, the following two measures are adopted in this embodiment to cope with this situation. First, the capacities of the feed water pumps 28, 29 are made large. The minimum capacities required of the feed water pumps 28, 29 are 2W.sub.FW /3 and W.sub.FW /3, as described above. If these pump capacities are made large, the scope of control of the reactor power and the scope of control of the reactor water level become large by that portion. Hence, the pump capacity of the feed water 28 is increased to 300%, the pump capacity of the feed water pump 29 to 105%, and the pump capacity of the condenser pump 23 to 170% of the conventional levels. As a result, even if a sudden decline occurs in the reactor water level L.sub.R, it is possible to temporarily increase the feed water sparger flow rate W.sub.S. Secondly, a measure is taken to provide protection by means of the plant status discriminator 73. The arrangement of the plant status discriminator 73 will be described with reference to FIG. 8. The plant status discriminator 73 has flow-rate request signal adjusters 73A, 73B, a limiter 73C, a status discriminator 73D, a temperature estimator 73E, an operating status discriminator 73F, and an adder 73G. Input to the plant status discriminator 73 are a trip signal SW.sub.T, a plant status amount X.sub.P measured by a detector installed in the BWR power plant, the feed water temperature T.sub.J in the feed water pipe 31 measured by the thermometer 54, the feed water temperature T.sub.S in the feed water pipe 33 measured by the thermometer 55, and the reactor power Q.sub.R measured by the power detector 56, as well as the sparger flow-rate request signal W.sub.SO and the jet pump driving water flow-rate request signal W.sub.JD. Upon receiving the trip signal SW.sub.T, the status discriminator 73D discriminates whether or not the plant status amount X.sub.P, which is subject to the effect of that trip signal SW.sub.T, has deviated from a predetermined level. When the trip signal SW.sub.T is not input and when the plant status amount X.sub.P is within the predetermined level even if the trip signal SW.sub.T has been input, the status discriminator 73D outputs as a valve changeover signal S.sub.9C an open signal to the shut-off valve 35 and a close signal to the shut-off valve 44. When it is judged that the plant status amount X.sub.P has deviated from the predetermined level, the status discriminator 73D outputs to at least one of the flow-rate request signal adjustors 73A, 73B at least one relevant signal of signals S.sub.S, S.sub.J for instructing a change in the flow-rate request signals in correspondence with the trip signal SW.sub.T. At the same time, the status discriminator 73D outputs the valve changeover signal S.sub.9C to the shut-off valves 35, 44 so as to close the shut-off valve 35 and open the shut-off valve 44 in correspondence with the trip signal SW.sub.T. No need arises to change over the opening and closing of the shut-off valves 35, 44 depending on the trip signal SW.sub.T. Upon receiving the change instruction signal S.sub.S, the flow-rate request signal adjuster 73A corrects the sparger flow-rate request signal W.sub.SO into the sparger flow-rate request signal W.sub.SP in response to the signal S.sub.S, and outputs the signal W.sub.SP to the feed water pump 29. Upon receiving the change instruction signal S.sub.J, the flow-rate request signal adjuster 73B corrects into the jet pump driving water flow-rate request signal W.sub.JD a jet pump driving water flow-rate request signal W.sub.JO, output from the limiter 73C in response to that signal S.sub.J, and outputs that signal W.sub.JD to the feed water pump 28. When a discrimination result signal LM output from the operating status discriminator 73F is zero (i.e., when the operating status is normal), the limiter 73C outputs to the flowrate request signal adjuster 73B the jet pump driving water flow-rate request signal W.sub.JO as it is as the jet pump driving water flow-rate request signal W.sub.JO'. A description will now be given of the operation of the flow-rate request signal adjusters 73A, 73B and the status discriminator 73D with respect to a specific trip signal SW.sub.T, e.g. a Main Steam Separator Valve Close signal. Upon issuance of the Main Steam Separator Valve Close signal, the control rods 4 are inserted rapidly into the core 2, with the result that the core 2 is set in a scram. When the reactor water level L.sub.R has dropped to a predetermined level, the status discriminator 73D, upon receiving the Main Steam Separator Valve Close signal, outputs the following signals: the change instruction signal S.sub.S of Feed water Sparger Flow-Rate Request Signal W.sub.SO increase, the change instruction signal S.sub.J of Jet Pump Driving Water Flow-rate Request Signal W.sub.JO, and the valve changeover signal S.sub.9C for closing the shut-off valve 35 and opening the shut-off valve 44. When the trip signal SW.sub.T is the Main Steam Separator Valve Close signal, the change instruction signals S.sub.S, S.sub.J and the valve changeover signal S.sub.9C are output. The shut-off valve 35 is closed and the shut-off valve 44 is opened by the valve changeover signal S.sub.9C. The flow-rate request signal adjuster 73A outputs the feed water sparger flowrate request signal W.sub.SP in which the feed water sparger flow-rate request signal W.sub.SO is increased on the basis of the change instruction signal S.sub.S. The flow-rate request signal adjuster 73B outputs the jet pump driving water flow-rate request signal W.sub.JO' as the jet pump driving water flow-rate request signal W.sub.JD on the basis of the change instruction signal S.sub.J. These signals do not change the number of revolutions of the feed water pump 28, but increases the number of revolutions of the feed water pump 29. The feed water delivered from the feed water pump 28 is supplied from the feed water sparger 8 into the reactor pressure vessel 1 via the feed water pipe 31, the bypass pipe 43, and the feed water pipe 33. Although the flow rate of the feed water supplied from the feed water sparger 8 increases, the reactor water level L.sub.R inside the reactor pressure vessel 1 is held at a predetermined level by virtue of the function of the feed water/reactor power controller 70, particularly the water level controller 71. Incidentally, the reason for closing the shut-off valve 35 and opening the shut-off valve 44 by means of the valve changeover signal S.sub.9C is to prevent a large quantity of the reactor flow from being supplied into the reactor core 2 after a scram as a result of driving the jet pumps 6 on the basis of the operation of the feed water delivered from the nozzles 32. When a turbine trip has occurred, the status discriminator 73D does not output the valve changeover signal S.sub.9C but outputs the change instruction signal S.sub.S and S.sub.J for decreasing the feed water sparger flow-rate request signal W.sub.SO and the jet pump driving water flow-rate request signal W.sub.JO' to a predetermined level. By virtue of the above-described arrangement, it is possible to prevent the reactor water level L.sub.R from undergoing a sudden change during tripping of various kinds. The status discriminator 73D has a memory for storing the change instruction signals S.sub.S and S.sub.J and the valve changeover signal S.sub.9C with respect to various kinds of trip signal SW.sub.T. This memory may be provided separately from the status discriminator 73D. The status discriminator 73D retrieves from the memory each signal corresponding to the trip signal SW.sub.T input. The plant status discriminator 73 has a first protecting section constituted by the flow-rate request signal adjusters 73A, 73B and the status discriminator 73D and adapted to correct each flow-rate request signal during a trip as well as a second protecting section for restricting the jet pump driving water flow-rate request signal when the operating status has deviated from the allowable range, which will be described below. A detailed description will now be made of the second protecting section. The temperature estimator 73E determines the temperature T.sub.JSC of the suction flow 45A sucked into the jet pumps 6 on the basis of the temperature of the feed water sparger water in the same way as the temperature estimator 46A. The temperature difference .DELTA.T between the temperature T.sub.JSC and the temperature T.sub.J of the jet pump driving water is determined by the adder 73G. Upon receiving the reactor power Q.sub.R and the temperature difference .DELTA.T, the operating status discriminator 73F discriminates whether or not the current operating status determined by these values is within an allowable range. The operating status discriminator 73F stores characteristics based on the reactor power Q.sub.R and the temperature difference .DELTA.T, as are shown in FIG. 10. In FIG. 10, the area above the line L.sub.1 is an operation prohibited area. The line L.sub.2 represents an alarm issuing line, and the area below the alarm issuing line L.sub.2 is the operation allowed area. When the position determined by the reactor power Q.sub.R and the temperature difference .DELTA.T is in the operation allowed area, the operating status of the BWR power is normal. When that position is in the area above the alarm issuing line L.sub.2, the operating status of the BWR power plant is in an abnormal status. The operating status discriminator 73F outputs the discrimination result signal LM with a level "0" when the operating status is normal, and outputs the discrimination result signal LM with a level "1" when the operating status is abnormal. When the operating status is abnormal, the point X determined by the reactor power Q.sub.R and the temperature difference .DELTA.T is located in the area above the alarm issuing line L.sub.2. When the point X has come into contact with the alarm issuing line L.sub.2 from below, the operating status discriminator 73F issues an alarm, and this alarm is displayed on a display unit (not shown). The discrimination result signal LM is input to the limiter 73C. The limiter 73C has the control characteristics shown in FIG. 11. The solid line represents the characteristics at the time when the discrimination result signal LM is "0", while the broken lines represent the characteristics at the time when the signal LM is " 1". Accordingly, the limiter 73C restricts the jet pump driving water flow-rate request signal W.sub.JO input when the operating status is abnormal, to the values of the jet pump driving water flow-rate request signal W.sub.JO, indicated by the broken line. As a result, the point X is held at a position on the alarm issuing line L.sub.2. Accordingly, the flow rate of the jet pump driving water is prevented from being controlled to such a flow rate which involves the occurrence of the cavitation in the jet pumps 6. (IV) Control for Improving the Response Characteristics of Generator Output This control is effected by the turbine controller 80 and the blending valve controller 90. A detailed arrangement of these controllers is shown in FIG. 12. The turbine controller 80 has a pressure controller 80A, a speed controller 80B, a low-value preference gate 80C, an initial pressure adjuster 80D, controllers 80E, 80F, and adders 80G-80J. The bleeding valve controller 90 has a bleeding valve controller 90A, a bleeding amount distributor 90B, a jet pump driving water flow-rate compensating/controlling device 90C, and adders 90D, 90E. The electric generating power of the generator 16 is determined by a real amount of steam flowing through the high- and low-pressure turbines 13, 14. This amount of steam is adjusted by the steam regulator valve 20. There are two modes in controlling the amount of opening of the steam regulator valve 20. The first mode is designed to control the steam pressure P.sub.S of the reactor pressure vessel 1 (measured by the pressure gauge 57 provided in the main steam pipe 17) such as to assume a value close to the set pressure P.sub.SO (the reactor pressure is controlled to a fixed level). Namely, the measured pressure P.sub.S, the set pressure P.sub.SO, and the initial pressure which is an output of the initial pressure adjuster 90D are added to the adder 80G, and a deviation signal between the initial pressure and the measured pressure P.sub.S on the one hand, and the set pressure P.sub.SO on the other, is input to the pressure controller 80A. The pressure controller 80A outputs a signal V.sub.1 on the basis of the deviation signal. The signal V.sub.1 is output to the low-value preference gate 80C and the adders 80I, 80H. The low-value preference gate 80C selects a lower-level signal between the signal V.sub.1 and a signal V.sub.2 and then outputs the same to the controller 80F. Normally, the signal V.sub.1 is smaller and is therefore selected. The controller 80F controls the amount of opening of the steam regulator valve 20 on the basis of the output signal of the low-value preference gate 80C. The controller 80E controls the amount of opening of the bypass valve 22 on the basis of the signal V.sub.1 obtained by the adder 80I as well as a signal of deviation between the output signal of the low-value preference gate 80C and a bias signal. For instance, if the reactor output increases, the steam pressure also increases, so that the steam regulator valve 20 is opened by the signal V.sub.1 so as to increase the amount of steam supplied to the turbine. Consequently, the output of the generator 5 increases. The second mode is designed to cope with requirements on the power system side as well as changes in the rotational speed of the generator 15. The speed controller 80E determines a value of deviation between a targeted value L set and the rotational speed R.sub.M of the generator 15 (turbine ) measured by the tachometer 58, and outputs the signal V.sub.2 corresponding to this deviation value. In addition, the speed controller 80B extracts from the signal V.sub.2 a component of a small output change in a short period (in a period of several tens of seconds and with the amplitude of variation being within 5%) which require speedy response, thereby prepares a load change request signal e , and outputs the same to the bleeding valve controller 90A. The signal V.sub.2 is input to the low-value preference gate 80C and the adder 80J. When not the signal V.sub.1 but the signal V.sub.2 is selected by the low-value preference gate 80C, the controller 80F effects the control of the amount of opening of the steam regulator valve 20 on the basis of the signal V.sub.2. The adder 80J outputs to the adder 80H a signal in which the signal V.sub.2 is added to the bias signal. The adder 80H, in turn, outputs a signal of deviation (a jet pump driving water flow-rate change request signal .epsilon..sub.Q) between the signal V.sub.1 and the output signal of the adder 80J to the initial pressure adjuster 80D and the adder 90E of the bleeding valve controller 90. The jet pump driving water flow-rate change request signal .epsilon..sub.Q is a reactor output change request signal. The above-described two modes are selected by the low-value preference gate 80C. A bias value of 10% in terms of the reactor power is added to the signal V.sub.2 by the speed controller 80B in such a manner that the first mode is selected at the time of the normal operation of the BWR power plant. However, when the variation of the rotational speed R.sub.M is large and the V.sub.2 fluctuates substantially, the steam regulator valve 20 is adjusted by the signal V.sub.2. In order to cause the generator output to respond speedily, the initial pressure regulator 80D is adapted to auxiliary move the steam regulator valve 20 before the appearance of the effect of the response (considerably lagging) of the water level/output controller 70 for controlling the flow rate of the jet pump driving water. In this example, the output signal of the bleeding valve controller 90A is used to control the bleeding valves 37, 39, and 41, and that output signal is also used to compensate the jet pump driving water flowrate change request signal .epsilon..sub.Q output from the turbine controller 80. The load change request signal .epsilon..sub.LOAD obtained by the compensation is output to the reactor power master controller 72 of the water level/reactor power controller 70 via the general controller 60. Upon receiving the load change request signal .epsilon..sub.L, the bleeding valve controller 90A determines a bleeding valve closing-amount signal .epsilon..sub.V (the opposite of the opening amount signal) on the basis of that signal .epsilon..sub.L, and outputs this signal to the bleeding amount distributor 90B and the jet pump driving water flow-rate compensation controller 90C. The bleeding amount distributor 90B designates each closing amount to the bleeding valves 37 and 39, and designates the amount of opening the bleeding valve 41 on the basis of the temperature difference .DELTA.T output from the feed water temperature controller 46. The control of the amount of opening of the bleeding valve 41 by the bleeding amount distributor 90B is performed when the temperature difference .DELTA.T has exceeded the alarm issuing line L.sub.2 shown in FIG. 10. The jet pump driving water flow-rate compensation controller 90C outputs a jet pump driving water flow-rate compensation signal .epsilon..sub.FW in proportion to the bleeding valve closing-amount signal .epsilon..sub.V input, so as to compensate a decline in the feed water temperature at the reactor inlet. This signal .epsilon..sub.FW is input to the adder 90E. The adder 90E effects compensation by adding the compensation signal .epsilon..sub.FW to the jet pump driving water flow-rate change request signal .epsilon..sub.Q, and outputs to the general controller 60 the load change request signal .epsilon..sub.LOAD obtained by compensation. Upon receiving the load change request signal .epsilon..sub.LOAD from the general controller 60, as described above, the water level/reactor power controller 70 adjusts the number of revolutions of the feed water pump 28 on the basis of this request signal. The core flow rate is thus altered. The arrangement of the bleeding valve controller 90A and the bleeding amount distributor 90A will now be described. Specifically, the bleeding valve controller 90A has the arrangement shown in FIG. 13 and is provided with a switch 90A.sub.1 and an adjuster 90A.sub.2. The switch 90A.sub.l is normally closed, and can be opened, for instance, manually when the functions of this embodiment are not performed. The adjuster 90A.sub.2 is a proportional/integration-type adjuster. As shown in FIG. 14, the bleeding amount distributor 90B has bleeding valve opening-amount controllers 91, 92 and a function generator 93. The function generator 93 is connected to the output detector 56 and the adder 9aA. The bleeding valve opening-amount controller 91 is adapted to control the amount of opening of the bleeding valve 41 and is provided with adder 91A, 91D, a function generator 91B, and a PI controller 91C. The adder 91 is also connected to the feed water temperature controller 46 and the function generator 91B. The PI controller 91C is connected to the function generator 91B. The adder 91D is connected to an output terminal of the PI controller 91C and an input terminal of the signal selector 47. The bleeding valve opening-amount controller 92 for adjusting the amounts of opening of the bleeding valves 37, 39 has adjusters 92A, 92B connected to the adder 90D, the adjuster 92A being connected to the bleeding valve 37, the adjuster 92B being connected to the bleeding valve 39. The response of each part according to this embodiment will now be described below. First, an increased load of the electric power system results in a decline in the rotational speed (detected by the tachometer 58) of the generator 15 (turbine). This decline is detected by the speed controller 80B as the load change request signal .epsilon..sub.L (normally zero; in this case, a positive value). Upon receipt of the load change request signal e.sub.L, the bleeding valve controller 90A determines the closing-amount signal .epsilon..sub.V which indicates an amount by which the bleeding valve should be closed by a proportional/integration-type adjuster 90A.sub.2 (with a proportional gain K.sub.P and an integral gain K.sub.I). This closing-amount signal .epsilon..sub.V is transmitted to the jet pump driving water flow-rate compensation controller 90C and the adder 90D. Since the function of the controller 90C has already been described, a description thereof will be omitted. A deviation signal V.sub.3 determined by the adder 90D is transmitted to the bleeding amount distributor 90B, and is input to the adjusters 92A, 92B of the bleeding valve opening-amount controller 92. The adjuster 92A output to the bleeding valve 37 an opening-amount signal ST.sub.3 corresponding to the deviation signal V.sub.3. The bleeding valve 37 is thus adjusted to a corresponding amount of opening on the basis of the opening-amount signal ST.sub.3. The adjuster 92B outputs to the bleeding valve 39 an opening-amount signal ST.sub.4 corresponding to the deviation signal V.sub.3. The bleeding valve 39 is thus adjusted to a corresponding amount of opening on the basis of the opening-amount signal ST.sub.4 The bleeding valve opening-amount controller 92 controls the amounts of opening of the bleeding valves 37, 39 so as to speedily respond to changes in the output of the generator 15 (in a period of several tens of seconds and the amplitude of variation falling within 5%) on the basis of the output of the tachometer 58. Therefore, it is possible to cope within a short time with variation in the output of the generator 15 which have short periods and small amplitude. In other words, since the amounts of opening of the bleeding valves 37, 39 are reduced sharply on the basis of the opening-amount signals ST.sub.3, ST.sub.4 to reduce the extracted amounts of steam, with the result that the rotational speeds of the turbines 13, 14 are increased and .epsilon..sub.L is returned quickly to zero. The temperature difference .DELTA.T obtained from the adder 46B of the feed water temperature controller 46 is input to the adder 9aA of the bleeding valve opening-amount controller 91. The function generator 93 determines the temperature difference .DELTA.T.sub.1 on the basis of the reactor power Q.sub.R measured by the reactor power detector 56. The function generator 93 stores a relation (equivalent to the formula of the alarm issuing line L.sub.2) indicating the relationships between the temperature difference .DELTA.T.sub.1 and the reactor power Q.sub.R The temperature difference .DELTA.T.sub.1 obtained by the function generator 93 is input to the adder 91A, which calculates the deviation signal U (=.DELTA.T.sub.l -.DELTA.T). When the deviation signal U is 0 or above, the function generator 91B outputs a signal V.sub.4 having a level 0, and outputs negative signal V.sub.4 corresponding to a case where the deviation signal U is negative. The PI controller 91C outputs a control signal on the basis of the signal V.sub.4. The adder 91D determines an opening-amount signal ST.sub.2 on the basis of this control signal, and outputs the opening-amount signal ST.sub.2 to the signal selector 47. As described above, the signal selector 47 selects either the opening-amount signal ST.sub.1 or ST.sub.2 thus input, and transmits the selected opening-amount signal to the bleeding valve 41. The amount of opening of the bleeding valve 41 is adjusted on the basis of the opening-amount signal selected by the signal selector 47. By decreasing the amount of opening of the bleeding valve 41 on the basis of the opening-amount signal ST.sub.2, it is possible to promptly shift the operation to that in the operation allowed area below the alarm issuing line L.sub.2 even if the temperature difference .DELTA.T corresponding to a given reactor output Q.sub.R exceeds the alarm issuing line L.sub.2. The bleeding valve opening-amount controller 91 is a kind of protector which is designed to shift the operation of the nuclear reactor to that in a safe status when the operation of the reactor is taking place in the area above the alarm issuing line L.sub.2. The response characteristics of control based on the function of the bleeding valve opening-amount controller 92 is shown in the time axis on a short-time scale in FIGS. 15A to 15G. The effect of an increase or decrease of the extracted amount of steam affects the turbine inlet side and the reactor pressure vessel side as well, and various variables follow a system output request QL (reflected on the load change request signal .epsilon..sub.L) accompanied by their peculiar delays and amplitude. Accordingly, it is possible to obtain extremely good load-following characteristics as a whole. However, when a situation occurs in which the load change request signal .epsilon..sub.L oscillates only on the positive value side over a relatively long period of time (more than several minutes) due to a change in the load requirements of the electric power system and a set point of load, the turbine output at this time also responds immediately by the control of the extracted steam described above. However, since the bleeding valves are held in a slightly closed state rather than in a fully closed state, the extracted amounts of steam supplied to the feed water heaters 25, 26 become small, so that the feed water temperatures being to decline gradually. In this embodiment, to compensate for this drop in the feed water temperatures, the load change request signal .epsilon..sub.LOAD obtained by the signal .epsilon..sub.FW proportional to the bleeding valve closing-amount signal .epsilon..sub.V is transmitted to the reactor power master controller 72 of the water level/reactor power controller 70 via the general controller 60, as described above. In other words, as shown in FIG. 13, since an integral term KI/S (KI denotes an integral gain) is included in the bleeding valve controller 90A, the signal .epsilon..sub.V indicates an integral value of the signal .epsilon..sub.L, or, to put it differently, an integral amount of the extracted amount of steam cut. Accordingly, if the flow rate of the jet pump driving water (core flow rate) is increased by a portion commensurate with this integral amount, the reactor power gradually bears the increment of the turbine output, so that the cut amount of the extracted steam finally becomes zero. Hence, a substantial decline in the feed water temperatures does not occur. To give a detailed description, as the flow rate of the jet pump driving water increases (whose speed is determined by the water level/reactor power controller 70), the reactor power increases, and the turbine controller 80 opens the steam adjusting valve CV to make the reactor power constant. Therefore, the rotation speed of the turbine (generator 15) tends to be increased further higher than the target value, the level of the signal becomes negative value, the level of the signal tends to be gradually lowered, and the bleeding valves 37, 39 tend to be more widely or more fully opened. In this embodiment, the increase in the output or power of the generator 15 at the time of increase in the set load in case of the operation following the load or at the time of increase in the load of the power system is initially compensated or supported by the cut of an amount of the bleeding steam, and is then compensated or supported by the increase in the reactor power. When the reactor power is increased, the opening-amounts or degrees of opening of the bleeding valves 37, 39 are returned to the original levels of 100 %, and the temperature of the feed water is not continued to be lowered. Especially, as the recirculating flow rate request signal .epsilon..sub.Q is adjusted according to the signal .epsilon..sub.FW, the decrease in the temperature of the feed water can be limited in a short time and the thermal shock to the structure in the reactor pressure vessel can be moderated. The response of various variables or parameters with respect to time in the long-term time scale are shown in FIGS. 15A through 15G. When the feed water temperature is lowered and when the increase/decrease in the temperature of the feed water is repeated at a considerable period, it is conceivable first and foremost that there are cases where the repeated thermal stress in the structural materials or structures such as the feed water nozzles may become a problem or the cause of the problem. In this embodiment, the thermal shock which occurs in internal structures at this time can be restrained, and the thermal fatigue occurring in the structures can be reduced remarkably. Secondly, it is conceivable that the core 2 experiences a large change in the thermal behavior. Namely, when the reactor power is increased by the same degree, an amount of shift in the boiling start point is small in the case where the increase in the reactor power is based on an increase in the core flow rate, whereas the boiling start point sifts substantially in the case where it is based on a decline in the feed water temperatures. Concomitantly, the axial distribution of output increases substantially in an average manner in the case where the reactor power is increased by an increase in the core flow rate. However, in the case where the reactor power is increased by the decline in the feed water temperature, local changes are large, and the strain becomes great as compared with the distribution prior to the change in the reactor output. Accordingly, changes in the feed water temperatures entail a shift in hot spots, and therefore the position where the hot spots occur shifts repeatedly. In accordance with this embodiment, this problem can also be solved. Furthermore, in this embodiment, the provision of the bleeding valve controller 90 enables changes in the outputs with small amplitude and short periods (normally, this is called an AFC operation and the governor-free operation mode). In addition, since the combined used is made of the control of the flow rate of the jet pump driving water (controlled by the water level/reactor power controller 70), which is the feed water for controlling reactor power changes with large and relatively slow fluctuations, the present invention provides a function of changing reactor power with large fluctuations. It should be noted that the bleeding valve opening-amount controller 92 and the function generator 93 may be removed from the bleeding amount distributor 90B and may be installed in the feed water temperature controller 46 together with the signal selector 47. (V) General Controller Lastly, a description will be given of a specific arrangement of the general controller 60. The general controller 60 controls the water level/reactor power controller 70, turbine controller 80, bleeding valve controller 90, and control rod driving apparatus controller 100, and transmits and receives signals to and from these controllers. The general controller 60 has a load change request signal evaluating section 60A and a controller selecting section 60B. Information concerning the status of the overall plant including the controllers is also input to the general controller 60. The load change request signal evaluating section 60A analyzes and evaluates the load change request signal .epsilon..sub.LOAD output from the bleeding valve controller 90. Namely, the load change request signal e.sub.LOAD is evaluated with respect to a variation amplitude .DELTA.Q and a time variation rate (or a cycle and a frequency component). The controller selecting section 60B has a controller selector 60C and adjusters 60D-60G. The controller selector 60C determines which signal should be transmitted to which controller with a certain degree of priority. For instance, the controller selector 60C outputs necessary control information, i.e., a component of a output change of a several percent in a period of several seconds to several minutes to the turbine controller 80, a component of an output change with small amplitude in a shorter period than the above period to the bleeding valve controller 90, a component of a reactor power change with large amplitude in a period of several minutes or more to the water level/reactor power controller 70, and a component of a slow, drifting output change to the control rod driving apparatus controller 100, respectively. In addition, information is exchanged among the adjusters 60D-60G , and adjustment is made by the controller selecting section 60B such that the overall controllers are free from mutual interference and thus become stable. For instance, the non-interference control is carried out on the basis of a theory on a multi-variable controller. A description will be given hereafter of the other embodiments of the present invention. Another embodiment provides a method of controlling the temperatures of the jet pump driving water and the feed water sparger water such as to assume targeted values by measuring these temperatures so that the cavitation will not occur during the operation of the jet pumps driven by the feed water. The arrangement in this case will be described with reference to FIG. 1. The extracted steam supplied to the feed water heaters 25, 26, and 27 is controlled by adjusting the bleeding valves 37, 38, and 41 respectively provided in the bleeding pipes 36, 38, and 40 on the basis of the temperatures (T.sub.J and T.sub.S) of the jet pump driving water and the feed water sparger water measured which are input to the bleeding valve controller 90. In other words, it is necessary for the bleeding valve controller 90 to operate the bleeding valves 37, 39 so as to control the temperature of the feed water sparger water, and also to adjust the amount of opening of the bleeding valve 41 by taking into account the amounts of opening of the bleeding valves 37, 39 so as to control the temperature of the jet pump driving water. The temperatures of the two kinds of feed water (those flowing through the feed water pipes 31 33) generally have a large time lag and a wasteful time, so that the bleeding valve controller 90 in this embodiment is preferably controlled on the basis of a program. Still another embodiment of the present invention is shown in FIG. 17. The nuclear power plant in accordance with this embodiment is provided with flow-rate control valves 61, 62 which substitute the feed pumps 28, 29 functioning as the feed water flow-rate controlling means in the feed water pipes 31, 33 in the embodiment shown in FIG. 1. Further, a feed water pump 63 is installed at the feed water pipe between the high-pressure feed water heater 26 and the branching point 34. The other arrangements of this embodiment are the same as those of the embodiment shown in FIG. 1. The flow-rate control valve 61 is disposed in the feed water pipe 31 between the high-pressure feed water heater 27 and the branching point 34. In addition, the flow-rate control valve 62 is disposed in the feed water pipe 33. The amount of opening of the flow-rate control valve 61 is controlled by the jet pump driving water flow-rate request signal W.sub.JD. The amount of opening of the flow-rate control valve 62 is controlled on the basis of the feed water sparger water flow-rate request signal W.sub.SP. With this embodiment, it is possible to obtain the same effect as that of the nuclear power plant shown in FIG. 1. FIG. 18 shows a further embodiment of the present invention. In this embodiment, a baffle cylinder 12 is added to the arrangement of the embodiment shown in FIG. 1. The baffle cylinder 12 is disposed between the riser portion 9 and the feed water sparger 8. The baffle cylinder 12 is disposed concentricaly with the riser portion 9, a lower end thereof extending down to the vicinity of the upper portions of the jet pumps 6. A multiplicity of small holes are provided in an upper portion of the baffle cylinder 12. A large quantity of high-temperature cooling water separated by the steam separator 10 is discharged outside the steam separator 10 and is mixed with the cooling water in an upper portion of the reactor pressure vessel 1. The temperature of this cooling water rises by being mixed with the large amount of high-temperature cooling water. The baffle cylinder 12 functions to control this high-temperature cooling water so as to be prevented from being mixed as a whole with the low-temperature feed water discharged from the feed water sparger 8. Since the baffle cylinder 12 is provided with the small holes, only a portion of the high-temperature cooling water is mixed with the low-temperature feed water. The high-temperature cooling water exists inside the baffle cylinder 12. Accordingly, the cooling water outside the baffle cylinder 12 maintains a state of low temperature and is sucked into the jet pumps 6. The nuclear power plant in accordance with this embodiment provides a similar effect to that of the embodiment shown in FIG. 1. Furthermore, by virtue of the effect derived from the installation of the baffle cylinder 12, the temperature of the cooling water sucked into the jet pumps 6 is lowered as compared to the embodiment shown in FIG. 1. FIGS. 19 and 20 show a still further embodiment of the present invention. These drawings illustrate a structure of a jet pump and its peripheral equipment in the reactor pressure vessel 1 in accordance with this embodiment. In this embodiment, the jet pump is arranged in two stages in a series. The other arrangements are identical to those of the embodiment shown in FIG. 1. Two jet pumps 6B, 6C are arranged in parallel. A jet pump 6A is disposed above the jet pumps 6B, 6C. The nozzle 32 is inserted into an upper end portion of the jet pump 6A. Two nozzles 64A, 64B are provided at a lower end portion of the jet pump 6A, i.e., the delivery side of the cooling water. These nozzles 64A, 64B are inserted into the upper end portions of the jet pumps 6B, 6C. If it is assumed that the M ratio of the jet pump 6A is M.sub.1 and that the M ratio of the jet pumps 6B, 6C is M.sub.2, the M ratio of the overall two-stage jet pumps is given by (M.sub.1 +M.sub.2 .times.(M.sub.1 +1)). A lower end of the baffle cylinder 12 disposed inside the feed water sparger 8 reaches an upper end of the jet pump 6A or thereabout. Reference numeral 12A denotes the small holes provided in the baffle cylinder 12. The jet pump 6A mainly sucks the low-temperature cooling water located outside the baffle cylinder 12 by means of the jet pump driving water (supplied through the feed water pipe 31) injected from the nozzle 32. The cooling water sucked into the jet pump 6A is injected from the nozzles 64A, 64B and serves as the driving water for the jet pumps 6B, 6C. This embodiment makes it possible to increase the M ratio without any occurrence of cavitation in the jet pumps. Moreover, the extent to which the occurrence of the cavitation can be prevented is large. A nuclear power plant in accordance with a further embodiment of the present invention will now be described with reference to FIG. 21. The same components and arrangements as those of the embodiment shown in FIG. 1 are denoted by the same reference numerals. This embodiment is provided with a feed water temperature controller 49, a water level/reactor power controller 75, and a bleeding valve controller 94 which substitute the feed water temperature controller 46, the water level/ reactor power controller 70, and the bleeding valve controller 90 in the embodiment shown in FIG. 1, and the signal selector 47 is eliminated. Furthermore, a thermometer 59 for measuring the temperature T.sub.SA of the cooling water (saturated water) is provided in the reactor pressure vessel 1 (e.g. in the riser portion). A pressure gauge 62 for measuring the reactor pressure P.sub.R is installed in the reactor pressure vessel 1. Measurement signals of the thermometer 59 and the pressure gauge 62 are input to the water level/reactor power controller 75. A description will now be made of the feed water temperature controller 49 with reference to FIG. 22. The feed water temperature controller 49 has a temperature estimator 46A, adders 46B, 49C, 49E, and 49G, function generators 49A, 49B, a targeted temperature difference setter 49D, and a PID controller 49F. The respective input terminals of the function generators 49A, 49B are connected to the output detector 56, while the respective output terminals thereof are connected to the adder 49C. The adder 49C is connected to the targeted temperature difference setter 49D. An input terminal of the adder 49E is connected to the targeted temperature difference setter 49D and the adder 46B, while an output terminal thereof is connected to the PID controller 49F. The PID controller 49F is connected to the bleeding valve 41 via the adder 49G. The function generator 49a stores characteristics L.sub.U2 (corresponding to the characteristics of an alarm issuing line L.sub.U2 stored in an operating status discriminator 76A shown in FIG. 24) indicating the relationships between the reactor power Q.sub.R and the temperature difference .rarw.T.sub.2. The function generator 49B stores characteristics L.sub.D2 (corresponding to an alarm issuing line L.sub.D2 which is stored in the operating status discriminator 76A and whose level is lower than that of the alarm issuing line L.sub.U2). The area between the alarm issuing line L.sub.U2 at the upper limit stored in the operation status discriminator 76A and the alarm issuing line L.sub.D2 at the lower limit represents the operation allowed area. The function generator 49A outputs the temperature difference .DELTA.T.sub.2 corresponding to the reactor power Q.sub.R input. The function generator 49B outputs the temperature difference .DELTA.T.sub.3 corresponding to the reactor power Q.sub.R input. The adder 49C adds the temperature differences .DELTA.T.sub.2 and .DELTA.T.sub.3 and outputs a temperature difference .DELTA.T.sub.4 thus obtained. The targeted temperature difference setter 49D calculates a targeted temperature difference .DELTA.T* on the basis of the temperature difference .DELTA.T.sub.4 input. In this embodiment, a temperature difference which is 0.5-fold the temperature difference .DELTA.T.sub.4, i.e., an average value of the temperature differences .DELTA.T.sub.2 and .DELTA.T.sub.3, constitutes the targeted temperature difference .DELTA.T. The adder 49E calculates a deviation between the temperature difference AT obtained by the calculator 46B and the targeted temperature difference .DELTA.T. The PID controller 49F outputs a control signal on the basis of this deviation in such a manner that the temperature of the jet pump driving water becomes higher than that of the feed water sparger water by a predetermined temperature. The adder 49G outputs an opening-amount signal ST.sub.5 on the basis of this control signal. The amount of opening of the bleeding valve 41 is controlled on the basis of the opening-amount signal ST.sub.5. In this embodiment since the feed water heater 27 and the feed water pipes 31, 33 are provided, it is possible to effect the above-described control mode of item (I) which allows a function similar to that of the embodiment shown in FIG. 1 to be obtained. The feed water temperature controller 49 controls the difference between the temperatures of the feed water flowing through the feed water pipe 31 and the feed water flowing through the feed water pipe 33 to a predetermined temperature. A description will now be made of the water level/reactor power controller 75 for carrying out the above-described control mode of item (II). FIG. 23 illustrates the arrangement of the water level/reactor power controller 75. This water level/reactor power controller 75 has the water level controller 71 and the reactor power master controller 72 both provided in the water level/reactor power controller 75. In addition to these, the water level/reactor power controller has a plant status discriminator 76. At the time of carrying out the control mode of item (II), the water level controller 71, the reactor power master controller 72, and the plant status discriminator 76 of the water level/ reactor power controller 75 operates in the same manner as that carried out by the water level controller 71, the reactor power master controller 72, and the plant status discriminator 73 of the water level/reactor power controller 75. In this embodiment, the plant status discriminator 76 is provided with the protective function performed by the plant status discriminator 73 in the control mode of item (III) in the embodiment shown in FIG. 1. A detailed arrangement of the plant status discriminator 76 is shown in FIG. 24. In the same way as the plant status discriminator 73, the plant status discriminator 76 has the flow-rate request signal adjusters 73A, 73B, the limiter 73C, the status discriminator 73D, the temperature estimator 73E, and the adder 73G. In addition to the foregoing arrangement, the plant status discriminator 76 has an operating status discriminator 76A, thermal shock discriminator 76B, 76C, and an OR circuit 76D. The thermal shock discriminator 76B is connected to the thermometers 54, 59 and the output detector 56. The thermal shock discriminator 76C is connected to the thermometers 55, 59 and the output detector 56. The respective output terminals of the operation status discriminator 76A and the thermal shock discriminators 76B, 76C are connected to the limiter 73C via the OR circuit 76D. Upon receiving the jet pump driving water temperature T.sub.J measured by the thermometer 54 and a saturated water temperature T.sub.SA measured by the thermometer 59, the thermal shock discriminator 76B effects discrimination in accordance with the following formula: EQU T.sub.SA -T.sub.J <.DELTA.T.sub.J (Q.sub.R) (1) The thermal shock discriminator 76B outputs "0" when Formula (1) is satisfied, and "1" when it is not satisfied. .DELTA.T.sub.J denotes an allowable maximum variation ratio of the jet pump driving water temperature which is capable preventing the occurrence of thermal shock and fatigue occurring as a result of supplying the jet pump driving water into the reactor pressure vessel 1, and is a function of the reactor power Q.sub.R. Upon receiving the feed water sparger water temperature T.sub.S measured by the thermometer 55 and the saturated water temperature T.sub.SA, the thermal shock discriminator 76C effects discrimination in accordance with the following formula: EQU T.sub.SA -T.sub.S <.DELTA.T.sub.S (Q.sub.R) (2) The thermal shock discriminator 76C outputs "0" when Formula (2) is satisfied, and "1" when it is not satisfied. .DELTA.T.sub.S denotes an allowable maximum variation ratio of the feed water sparger water temperature which is capable of preventing the occurrence of thermal shock and fatigue occurring as a result of supplying the feed water sparger water into the reactor pressure vessel 1, and is a function of the reactor power Q.sub.R. The operation status discriminator 76A stores the respective characteristics of the alarm issuing line L.sub.U2 indicating an upper limit of the temperature difference .DELTA.T and the alarm issuing line L.sub.D2 indicating a lower limit of the temperature difference .DELTA.T. The temperature difference .DELTA.T determined by these alarm issuing lines L.sub.U2, L.sub.D2 is a function of the reactor output Q.sub.R. The alarm issuing line L.sub.U2 corresponds to the alarm issuing line L.sub.2 shown in FIG. 10. The alarm issuing line L.sub.D2 is set such as to be lower than the alarm issuing line L.sub.U2 so as to prevent the occurrence of cavitation in the jet pumps 6. A line L.sub.U1 shown in FIG. 24 corresponds to the line L.sub.1 (FIG. 10). A line L.sub.D1 which is set below the alarm issuing line L.sub.D2 and indicates a temperature difference at a lower limit at which the cavitation does not occur in the jet pumps 6. The area above the line L.sub.U1 and the area below the line L.sub.D1 are the reactor operation prohibited areas. The area between the alarm issuing lines L.sub.U2 and L.sub.D2 is the operation allowed area. Upon receiving the temperature difference .DELTA.T output from the adder 73G, the operation status discriminator 76A determines whether or not the temperature difference .DELTA.T is present in the area between the alarm issuing lines L.sub.U2 and L.sub.D2 by reflecting the reactor power Q.sub.R. When YES is the answer, the operation status discriminator 76A outputs "0", and when NO is the answer, it outputs "1". Upon receiving the outputs of the thermal shock discriminators 76B, 76C and the operation status discriminator 76A, the OR circuit outputs the discrimination result signal LM of "1" or "0" to the limiter 73C. The protective function of the plant status discriminator 76 described above is the function of the second protecting section shown in the control mode in item (III) in the embodiment shown in FIG. 1. The function of the first protecting section in the plant status discriminator 76 in accordance with this embodiment can be attained by the flow-rate request signal adjusters 73A, 73B, the limiter 73C, and the status discriminator 73D, in the same way as the first protecting section in the embodiment shown in FIG. 1. The first and second protecting sections of the plant status discriminator 76 produce the same effect as that obtained by the first and second protecting sections of the plant status discriminator 73. Furthermore, since this embodiment is provided with the thermal shock discriminators 76B, 76C, it is possible to prevent the thermal shock and fatigue based on changes in the feed water temperatures. In addition, since the operation status discriminator 76A determines whether the temperature difference .DELTA.T is in the area above the alarm issuing line L.sub.D2 or in the area therebelow, and the limiter 73C is operated by the discrimination result signal LM based on this judgement, it is possible to prevent the occurrence of the cavitation in the jet pumps. The control mode of item (IV) carried out in the embodiment shown in FIG. 1 is performed by the bleeding valve controller 94 in this embodiment. As shown in FIG. 25, the bleeding valve controller 94 has the bleeding valve controller 90A, the jet pump driving water flow-rate compensation controller 90C, the adders 90D, 90E, and a bleeding amount distributor 95. The arrangements other than that of the bleeding amount distributor 95 are the same as those provided in the bleeding valve controller 90. The bleeding valve distributor 95 has a bleeding valve controller 96 and a bleeding valve opening-amount compensator 97, as shown in FIG. 26. The bleeding valve controller 96 has adjusters 92A, 92B, compensating amount distributors 96A, 96B, and adders 96C, 96D. An input terminal of the adder 96C is connected to the adjuster 92A and the compensating amount distributor 96A, while an output terminal thereof is connected to the bleeding valve 37. An input terminal of the adder 96D is connected to the adjuster 92B and the compensating amount distributors 96B, while an output terminal thereof is connected to the bleeding valve 39. The bleeding valve opening-amount compensator 97 has a feed water temperature target setter 97A to which the reactor power Q.sub.R is input, an adder 97B connected to the feed water temperature target setter 97A and the thermometer 55, a dead zone device 97C connected to the adder 97B, and a PID controller 97D connected to the dead zone device 97C and adapted to perform a proportional and integral operation. An output terminal of the PID controller 97D is connected to the compensating amount distributors 96A, 96B. The feed water temperature target setter 97A determines a targeted control value T.sub.S * of the feed water sparger water temperature T.sub.S on the basis of the reactor power Q.sub.R input. The adder 97B calculates a deviation (T.sub.S -T.sub.S ) between the feed water sparger water temperature T.sub.S and the targeted control value T.sub.S . The dead zone device 97 has dead zones with an amplitude of .delta. on the+side and the--side symmetrically about a zero line, and output a positive value when the deviation (T.sub.S -T.sub.S *) has become grater than .delta. and a negative signal when it has become smaller than -.delta.. The PID controller 97D outputs a compensation signal SA on the basis of an output signal of the dead zone device 97C. The compensation signal SA is input to the compensating amount distributors 96A, 96B. The adjusters 92A, 92B output the opening-amount signals ST.sub.3, ST.sub.4 on the basis of the deviation signal V.sub.3, as described above. The compensating amount distributor 96A calculates .alpha..SA, where .alpha. is a coefficient of distribution and assumes a value given by 0.ltoreq..alpha.<1. The compensating amount distributor 96B calculates (1-.alpha.).SA. The adder 96C adds .alpha..SA output from the compensating amount distributor 96A to the opening-amount signal ST.sub.3 to obtain an opening-amount signal ST.sub.6 , and outputs this signal ST.sub.6 to the bleeding valve 37. The amount of opening of the bleeding valve 37 is adjusted to an amount corresponding to the opening-amount signal ST.sub.6. The adder 96D adds (1-.alpha.).SA output from the compensating amount distributor 96B to the opening-amount signal ST.sub.3 to obtain an opening-amount signal ST.sub.7, and outputs this opening-amount signal ST.sub.7 to the bleeding valve 37. An amount of opening of the bleeding valve 37 is adjusted to an amount corresponding to the opening-amount signal ST.sub.7. The bleeding valves 37, 39 thus controlled operate on the basis of both a component of relatively slow variation which is included in the load change request signal .epsilon..sub.L and a component of variation with small amplitude and a short period which is included in the load change request signal .epsilon..sub.L. In this embodiment, it is possible to obtain an effect similar to that obtained in the control mode of item (IV) of the embodiment shown in FIG. 1. Since this embodiment is provided with the bleeding valve opening-amount compensator 97, the accuracy of controlling the amounts of opening of the bleeding valves 37, 39 improves as compared with those of the embodiment shown in FIG. 1. This embodiment is also provided with the general controller 60 having the same function as that of the embodiment shown in FIG. 1. Accordingly, the effect brought about by the general controller 60 shown in FIG. 16 can also be obtained in this embodiment.
	description	1. Field This invention relates generally to maintenance of under vessel components of a nuclear reactor and more particularly to an automated high speed under vessel work platform for servicing control rod drives and other under vessel components of a boiling water reactor. 2. Related Art Control rod drives are used to position control rods in boiling water reactors to control the fission rate and fission density, and to provide adequate excess negative reactivity to shut down the reactor from any normal operating or accident condition at the most reactive time in the core life. FIG. 1 is a sectional view of a boiling water nuclear reactor pressure vessel 10 with parts cut away to expose the interior thereof and illustrates the environment in which a control rod drive operates. The reactor pressure vessel 10 has a generally cylindrical shape and is closed at one end by a fixed bottom head 12 and at its other end by a removable head 14. A sidewall 16 extends from the bottom head 12 to the top head 14. A cylindrically-shaped core shroud 20 surrounds a reactor core 22. The shroud 20 is supported at one end by a shroud support 24 and includes a removable shroud head 26 at the other end. An annulus 28 is formed between the shroud 20 and the sidewall of the vessel 16. Heat is generated within the core 22, which includes fuel bundles 36 of fissionable material. Water circulated up through the core 22 is at least partially converted to steam. Steam separators 38 separate steam from water, which is recirculated. Residual water is removed from the steam by steam dryers 40. The steam exits the pressure vessel 10 through a steam outlet 42 near the vessel top head 14 and is commonly used to drive a turbine generator for the production of electricity. The fuel bundles 36 are aligned by a lower core plate 50 located at the base of the core 22. A top guide 52 aligns the fuel bundles 36 as they are lowered into the core 22. Core plate 50 and top guide 52 are supported by the core shroud 20. The amount of heat generated in the core 22 is regulated by inserting and withdrawing control rods 44 of neutron absorbing material, such as Hafnium. The control rods extend from underneath the vessel to within the core. To the extent that the control rods 44 are inserted between the fuel bundles 36, the control rods absorb neutrons that would otherwise be available to promote the chain reaction which generates heat in the core 22. The control rod guide tubes 46 below the lower core plate 50, align the vertical motion of the control rods 44 during insertion and withdrawal. Hydraulic control rod drives 48, which extend through the bottom head 12, effect the insertion and withdrawal of the control rods 44. Each control rod drive is mounted vertically in a control rod drive housing 18 which is welded to a stud tube which, in turn, is welded to the bottom head of the reactor vessel 12. A control rod drive flange is bolted and sealed to the flange of the control rod drive housing 18, which contains ports for attaching the control rod drive hydraulic system lines. Demineralized water supplied by the control rod drive hydraulic system serves as the hydraulic fluid for effecting control rod drive operation. Control rod drives are frequently removed from the reactor for servicing in order to maintain their reliability. All boiling water reactors use the same under vessel platform that was installed during original construction for servicing the control rods. The existing platforms are slow and can only be controlled by a technician under the vessel while on the platform using a hand pendant running to an air motor or electric AC motor. The only alternative is via a manually operated hand wheel. Besides being slow, the existing work platforms require under vessel technicians to take their mind off their tasks, in order to properly position a platform often causing setup under the wrong location. Due to this condition, and the operational process of the existing work platforms, many errors, close calls and equipment damage take place during planned outages. Accordingly, a new under vessel work environment is desired that will facilitate remote operation, new service techniques and a more efficient service operating environment. These and other objects are achieved by an under vessel automated work platform assembly for remotely servicing a lower portion of the vessel. The platform assembly includes a horizontal, generally circular work platform extending in a first plane and a generally circular rail extending in a second plane, substantially parallel to the first plane, with the rail supporting an orbital track on which the work platform is rotationally supported to rotate the work platform in the first plane. A remotely controlled motor rotates the work platform on the orbital track and a linear track extends across the diameter of the work platform. A carriage is remotely movable on the linear track across the diameter of the work platform. A robot attaches to the carriage and is moveable therewith. The robot has a vertically extending member moveable in a direction perpendicular to the first plane and has a receptacle proximate a distal end of the vertically extending member for supporting a tool. Preferably, the under vessel automated work platform assembly includes a hand wheel for manually moving the work platform around the orbital track as an alternative to the remotely controlled motor. In one embodiment, the robot is a swappable task robot and the work platform supports a camera generally focused on a distal end of the vertically extending member on the robot. Preferably, the camera is configured to have a remote controlled panning capability. In another embodiment, where the vessel is a nuclear reactor vessel having a nuclear core, the remotely controlled motor for moving the work platform on the automated work platform assembly and the remotely controlled carriage are configured to communicate with a controller and are responsive to coordinates of the reactor core inputted to the controller to move the vertically extending member under a core position associated with the coordinates. Preferably, the nuclear reactor vessel includes a refueling bridge configured to be positioned over the nuclear core and move one or more pieces of the refueling equipment over the nuclear core to directed ones of the coordinates inputted at a refueling system command station, wherein the controller and the refueling system command station automatically coordinate so that the refueling bridge and the controller do not address the same coordinates at the same time. The automated high speed under vessel work platform of this invention will be used to perform boiling water reactor under vessel maintenance and repair work remotely, without the need of personnel under the vessel. The system will perform these work activities more precisely, more accurately and faster, as well as offer concurrent verification at the time of the activity via the use of on-board high definition cameras. The automated high speed under vessel work platform has three automated axes consisting of a 360° rotational horizontal axis, a linear horizontal axis via a traversing trolley (also referred to as a carriage) and a linear vertical axis via swappable task robots which can also contain fourth and fifth axes capabilities. The swappable task robots may be attached by personnel or stored under the vessel and retrieved by the remotely operated horizontal trolley. Another advantage of the automated high speed under vessel work platform is the ability to automatically drive to a specific location merely by typing the core location into the control software. The automated high speed under vessel work platform has the capability to also be operated manually via a hand wheel and locally via a handheld pendant. The task robots will perform tasks such as shootout steel inspection, all other visual inspections, tag hanging, concurrent verifications, position indication probe removal, control rod drive uncoupling, control rod drive de-torqueing, control rod drive re-torqueing, and local power range monitor disengagement, local power range monitor drain can installation removal and control as well as control rod drive exchange. FIG. 2 shows a perspective view of the reactor previously shown in FIG. 1 with the under vessel work platform 30 supported below the control rod drives 18. The work platform 30 can be better appreciated from the views shown in FIGS. 3-8. The under vessel work platform assembly 30 has an orbital track 34 attached circumferentially to under vessel track mount locations 35. The orbital track 34 has a circumferential drive ring 54 having gear teeth 56. The work platform 32 is attached to the orbital track 34 via precision positive captive trucks 58 which can be best observed in FIGS. 7 and 8. The captive trucks allow for load transfers to the work platform 32 in any direction. The work platform 32 has an electrically powered rotational drive 60 comprising a pin gear which interfaces with the gear teeth 56 of the circumferential drive ring 54 allowing for 360° platform rotation. The work platform 32 has a linear track system 62 comprising a horizontal automated trolley 64 as can best be appreciated from FIGS. 4-8. The horizontal automated trolley 64 has an electrically powered drive 66 comprising a pin gear which interfaces with a gear rack 68 horizontally mounted within the work platform 32. With the circumferential axis and the horizontal axis, the horizontal automated trolley 64 can be remotely positioned at any location under the vessel. The automated horizontal trolley 64 has a remotely operated high definition camera 70 and spring-loaded power and position feedback pins 72. The pins 72 provide power to and feedback from swappable automated task robots 74 which are designed to be inserted into the automated horizontal trolley 64 as illustrated in FIG. 5. The automated task robots 74 perform a variety of work tasks on under vessel components as well as simple tasks such as hanging tags, operating valves, and concurrent verifications via visual observation through the camera 70. As can be seen in FIG. 13, this entire system can be operated locally through a pendant 76, at a contractor control point with a controller 84 or remotely through the control room 78. The controller 84 communicates with a refueling system command station 86 on the refueling floor 82 so the refueling bridge 88 and the under vessel work platform 32 are each aware of each other's location relative to the core so the proper work processes can be adhere to. Preferably the refueling floor 82 monitors the position of the trolley 64 on the work platform 32, but has no control over the positioning of the trolley. The control room 78 can control the cameras and monitor the position of the work platform 32. The controller 84 has both a view of the camera feed as well as complete control of the work platform assembly 30 and the pendant 76 can control the entire movement of the work platform assembly 30. As figuratively illustrated in FIG. 9, by typing a cell location into the software, the automated high speed under vessel work platform can automatically move to that cell location; eliminating the HuP (human performance) trap of working on the wrong location. FIGS. 10, 11 and 12 show three different swappable task robots which can be used with the horizontal automated trolley 64. The robot 74 in FIG. 10 has a third, fourth and fifth axes of movement. The robot 74 illustrated in FIG. 11 adds a third axis with a vertical rotational fourth axis 90. The robot 74 illustrated in FIG. 12 adds a third and fourth axes. The axes of movement of the robots illustrated in FIGS. 10, 11 and 12 are figuratively shown by the directional arrows superimposed on those figures. Accordingly, the under vessel work platform of this invention can be remotely operated and provide feedback to the operator of its exact location, while also providing live streaming video of the entire under vessel work area. The under vessel work platform can provide remote human inspections and scans of the entire under vessel area via the control room or other control point without the need of setup by personnel; inspections could even take place during normal plant operation, a feature currently not available at any boiling water reactor facility. The work platform has the ability to communicate with the refueling bridge, identifying the work location of each to their operators. The work platform of this invention can drastically cut the proposed under vessel radiation dose expectations for plant outages. The platform retains the capability of manual operation through a hand wheel that can be inserted at 80 shown in FIG. 5. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
	description	This application is a continuation application of International Application No. PCT/JP2004/007410 filed May 24, 2004, which is hereby incorporated by reference. This is a continuation from PCT International Application No. PCT/JP2004/007410 filed on May 24, 2004, which is hereby incorporated by reference The present invention relates to a multilayer film reflective mirror that is used in the X-ray optical system of an X-ray microscope, X-ray analysis device, X-ray exposure apparatus, or the like, and an X-ray exposure apparatus using this multilayer film reflective mirror. In recent years, as semiconductor integrated circuits have become finer, projection lithographic techniques using X-rays (which have a shorter wavelength (11 to 14 nm) than conventional ultraviolet light) instead of ultraviolet light have been developed in order to improve the resolving power of optical systems (which is limited by the diffraction limit of light) (for example, see D. Tichenor, et al., “SPIE,” 1995, Vol. 2437, p. 292). Such techniques have recently acquired the name of EUV (extreme ultraviolet) lithography, and have shown promise as techniques for obtaining a resolving power of 70 nm or finer, which cannot be realized with conventional photolithography using light rays with a wavelength of 190 nm. The complex refractive index n of a substance in the X-ray wavelength region is expressed by n=1−δ−ik (δ and k are real numbers, and i is a symbol indicating complexity). The imaginary part k of this refractive index expresses X-ray absorption. Since δ and k are extremely small compared to 1, the refractive index in this region is extremely close to 1. Accordingly, conventional transmissive refraction type optical elements such as lenses cannot be used, and optical systems utilizing reflection are used instead. In the case of oblique-incidence optical systems that utilize total reflection to reflect X-rays that are incident on the reflective surface from an inclined direction, the reflectivity is extremely small at angles of incidence that are smaller than the critical angle θc of total reflection (approximately 20° or smaller at a wavelength of 10 nm) (i.e., angle of incidence that are close to perpendicular). Here, furthermore, the angle of incidence refers to the angle formed by the normal of the plane of incidence and the optical axis of the incident light. Accordingly, multilayer film reflective mirrors are used in which numerous reflective surfaces (several tens to several hundreds of layers in one example) are formed by laminating a substance having an interfacial amplitude reflectivity that is as high as possible, and the thicknesses of the respective layers are adjusted on the basis of light interference theory so that the phases of the respective reflected waves match. These multilayer film reflective mirrors are formed by alternately laminating, on the surface of a substrate, substances in which the difference between the refractive index in the X-ray wavelength region used and the refractive index in a vacuum (=1) is large,. and substances in which this difference is small. Furthermore, since these multilayer film reflective mirrors can also reflect X-rays that are perpendicularly incident, an optical system can be constructed which has a smaller aberration than an oblique-incidence optical system using total reflection. Moreover, such multilayer film reflective mirrors have a wavelength dependence which is such that X-rays are strongly reflected in cases where Bragg equation2d sinθ=nλ(d: periodic length of multilayer film, θ: angle of incidence, λ: wavelength of X-rays) is satisfied. Therefore, various factors must be selected so that this equation is satisfied. Known examples of multilayer films used in multilayer film reflective mirrors include films using a combination of W/C multilayer films in which tungsten (W) and carbon (C) are alternately laminated, Mo/C multilayer films in which molybdenum (Mo) and carbon are alternately laminated, and the like. Furthermore, these multilayer films are formed by thin film forming techniques such as sputtering, vacuum evaporation and CVD (chemical vapor deposition). Moreover, if an Mo/Si multilayer film in which molybdenum (Mo) layers and silicon (Si) layers are alternately laminated is used in the wavelength region in the vicinity of 13.4 nm, a reflectivity of 67.5% can be obtained in the case of perpendicular incidence (angle of incidence: 0°), and if an Mo/Be multilayer film in which Mo layers and beryllium (Be) layers are alternately laminated is used in the wavelength region in the vicinity of 11.3 nm, a reflectivity of 70.2% can be obtained in the case of perpendicular incidence (for example, see C. Montcalm, “Proceedings of SPIE,” 1998, Vol. 3331, p. 42). Reflective mirrors using such multilayer films are also applied to reduction projection lithographic technology using soft X-rays that is referred to as EUVL (extreme ultraviolet lithography). FIG. 3 is a sectional view showing in model form the structure of a multilayer film reflective mirror used in conventional EUVL. This multilayer film reflective mirror 41 is a mirror in which an Mo/Si multilayer film 45 is formed on a substrate 43. In this Mo/Si multilayer film 45, an Mo layer 47 and an Si layer 49 are taken as one layer pair, and approximately 40 to 50 such layer pairs are laminated. The periodic length of this Mo/Si multilayer film 45 (i.e., the thickness of one layer pair) is approximately 7 nm, and the ratio (Γ) of the thickness of one Mo layer to the periodic length is approximately 0.35 to 0.4. Furthermore, the surface (upper surface in the figure) of the substrate 43 ordinarily has a concave shape; however, in order to simplify the description, a portion of the multilayer film reflective mirror is made horizontal in the figure, and the number of laminated layers is abbreviated. Incidentally, the multilayer film reflective mirror 41 is manufactured by sputtering (ion beam sputtering, magnetron sputtering, or the like), electron beam deposition, or the like; here, the high-reflectivity Mo/Si multilayer film 45 generally has a compressive internal stress of approximately −350 MPa to −450 MPa. As a result, the following problem arises: namely, the substrate 43 of the multilayer film reflective mirror 41 is caused to undergo deformation by the compressive internal stress of the Mo/Si multilayer film 45, so that wavefront aberration is generated in the optical system, thus causing a deterioration in the optical characteristics. Accordingly, in order to reduce the compressive stress of a multilayer film with a high X-ray reflectivity, a technique has been reported in which a first multilayer film is formed on a substrate, and a multilayer film (second multilayer film) with a high X-ray reflectivity is formed on top of this first multilayer film, so that the stress of the multilayer film reflective mirror as a whole is reduced (for example, see E. Zoethout, et al., “SPIE Proceedings,” 2003, Vol. 5037, p. 872, and M. Shiraishi, et al., “SPIE Proceedings,” 2003, Vol. 5037, p, 249). Here, the periodic length of the first multilayer film is substantially the same as the periodic length of the second multilayer film, so that Γ is comparatively large (e.g., Γ=0.7). Since such a first multilayer film has a tensile stress, the compressive stress of the second multilayer film can be reduced. Conventional stress reduction techniques will be described with reference to FIGS. 4 and 5. FIG. 4 is a diagram showing the stress in a multilayer film with respect to Γ in a case where an Mo/Si multilayer film with a periodic length of 7.2 nm and a laminated layer number of 50 layer pairs was formed by sputtering with Γ varied. In FIG. 4, the horizontal axis expresses Γ (−), which is the ratio of the thickness of one Mo layer to the periodic length. Furthermore, Γ=0 indicates an Si single-layer film with a thickness of 250 nm, and Γ=1 indicates an Mo single-layer film with a thickness of 250 nm. Moreover, in FIG. 4, the vertical axis expresses the stress (MPa) of the film, with negative values indicating a compressive stress, and positive values indicating a tensile stress. The stress of the Mo/Si multilayer film varies according to Γ; it is seen that in the range in which Γ is smaller than approximately 0.5, the stress is a compressive stress, while in the range in which Γ is larger than approximately 0.5, the stress is a tensile stress. As was described above, since Γ of the second multilayer film which has a high reflectivity is approximately 0.35 to 0.4, this film has a compressive stress of approximately −350 MPa to −450 MPa. On the other hand, by using a multilayer film having a Γ value that is greater than approximately 0.5 as the first multilayer film, it is possible to generate a tensile stress in the first multilayer film. Accordingly, the internal stress of the multilayer film as whole can be reduced by combining a second multilayer film that has a compressive stress and a first multilayer film that has a tensile stress. FIG. 5 is a sectional view which shows the structure of a conventional low-stress multilayer film reflective mirror in model form. In this multilayer film reflective mirror 51, a first multilayer film 57 is formed between a substrate 53 and a second multilayer film 55. The second multilayer film 55 is an Mo/Si multilayer film consisting of Mo layers 551 and Si layers 553; in this film, the periodic length is set at 7.2 nm, Γ is set at 0.35, and the number of laminated layers is set at 50 layer pairs, so that a high X-ray reflectivity can be obtained. On the other hand, the first multilayer film 57 is an Mo/Si multilayer film consisting of Mo layers 571 and Si layers 573, with the periodic length set at 7.2 nm, Γ set at 0.7, and the number of laminated layer set at 30 layer pairs. Furthermore, in order to simplify the description, a portion of the multilayer film reflective mirror is made horizontal in the figure, and the number of laminated layers is abbreviated. In this multilayer film reflective mirror 51, the second multilayer film 55 has a Γ value of 0.35, and therefore has a compressive stress, while the first multilayer film 57 has a Γ value of 0.7, and therefore has a tensile stress. Accordingly, the internal stress of the multilayer film as a whole can be reduced. However, when multilayer film reflective mirrors are actually manufactured using such a conventional stress reduction technique, the following problem is encountered: namely, although the internal stress of the multilayer films is reduced, the X-ray reflectivity drops. The present invention was devised in the light of such problems; it is an object of the present invention to provide a multilayer film reflective mirror with low internal stress in which the drop in reflectivity is suppressed, and to provide an X-ray exposure apparatus or the like using this multilayer film reflective mirror. The first invention that is used to achieve the object described above is a multilayer film reflective mirror comprising a first multilayer film which is formed by alternately laminating layers consisting of a substance (first substance) in which the difference between the refractive index of this substance in the soft X-ray region and the refractive index of a vacuum is large, and layers consisting of a substance (second substance) in which this difference is small, on a substrate, and a second multilayer film which is formed on top of the first multilayer film, and which is formed by alternately laminating layers consisting of the first substance and layers consisting of the second substance, wherein the thickness of the layers consisting of the first substance in the first multilayer film is substantially equal to or smaller than the thickness of the layers consisting of the first substance in the second multilayer film, and the ratio of the thickness of the layers consisting of the first substance to the thickness of the layers consisting of the second substance in the first multilayer film is different from the ratio of these thicknesses in the second multilayer film. In the multilayer film reflective mirror of the present invention, the thickness of the layers of the first substance in the first multilayer film and the thickness of the layers of the first substance in the second multilayer film are made substantially equal, or the thickness of the layers of the first substance in the first multilayer film is made smaller than the thickness of the layers of the first substance in the second multilayer film. As a result, an increase in the surface roughness caused by micro-crystallization of the layers of the first substance can be suppressed, so that a drop in the reflectivity of the multilayer film reflective mirror can be suppressed. Furthermore, by making the ratio of the thickness of the layers of the first substance to the thickness of the layers of the second substance different in the first multilayer film and second multilayer film, it is possible to reduce the internal stress of the second multilayer film by means of the internal stress of the first multilayer film. Accordingly, a multilayer film reflective mirror with a low internal stress in which any drop in reflectivity is suppressed can be obtained. Moreover, the statement that “the thicknesses are substantially equal” means that cases in which these thicknesses differ somewhat are also included in the scope of the present invention, as long as the difference in thickness is in a range which is such that this difference has no effect on the reflectivity of the multilayer film reflective mirror. In the multilayer film reflective mirror described above, it is desirable that the thickness of the layers consisting of the first substance in the first multilayer film be 50% to 120% of the thickness of the layers consisting of the first substance in the second multilayer film. As a result, the multilayer film can easily be formed, and the surface roughness can be securely kept to a permissible value or less, so that the effect on the reflectivity can be reduced. In the present invention, it is desirable that the first multilayer film have an internal stress that counteracts the internal stress of the second multilayer film. As a result, the internal stress of the second multilayer film can be more securely reduced by means of the internal stress of the first multilayer film. In the present invention, it is desirable that the periodic length be the total of the thickness of the layers of the first substance and the thickness of the layers of the second substance; furthermore, where Γ is the ratio of the thickness of the layers of the first substance to the periodic length, it is desirable that the Γ value of the first multilayer film be larger than the Γ value of the second multilayer film. Generally, the multilayer film has a compressive stress when the ratio (Γ) of the thickness of the layers of the first substance to the periodic length is small, and has a tensile stress when Γ is large. The Γ value of the second multilayer film is set at a small value in order to increase the X-ray reflectivity, so that the second multilayer film has a compressive stress. Accordingly, the Γ value of the first multilayer film is set at a large value so that the first multilayer film has a tensile stress, thus making it possible to reduce the compressive stress of the second multilayer film. In the present invention, it is desirable that the first substance be molybdenum (Mo). Furthermore, it is desirable that the second substance be silicon (Si). As a result, it is possible to obtain a multilayer film reflective mirror that is inexpensive and superior in terms of durability, and that has a high X-ray reflectivity. The second invention that is used to achieve the object described above is a multilayer film reflective mirror comprising a first multilayer film which is formed by alternately laminating molybdenum (Mo) layers and silicon (Si) layers on a substrate, and a second multilayer film which is formed on top of this first multilayer film, and which is formed by alternately laminating Mo layers and Si layers, wherein the thickness of the Mo layers of the first multilayer film is 1.2 nm to 3 nm, and the ratio of the thickness of the Mo layers to the thickness of the Si layers in the first multilayer film differs from the ratio of the thickness of the Mo layers to the thickness of the Si layers in the second multilayer film. In the multilayer film reflective mirror of the present invention, since an Mo/Si multilayer film is used as the multilayer film, it is possible to obtain a multilayer film reflective mirror which is inexpensive and superior in terms of durability, and which has a high X-ray reflectivity. Furthermore, since the thickness of the Mo layers in the first multilayer film is set at 1.2 nm to 3 nm, an increase in the surface roughness caused by micro-crystallization of the Mo layers can be suppressed, so that a drop in the reflectivity of the multilayer film reflective mirror can be suppressed. Moreover, since the ratio of the thickness of the Mo layers to the thickness of the Si layers is varied between the first multilayer film and second multilayer film, the internal stress of the second multilayer film can be reduced by means of the internal stress of the first multilayer film. Accordingly, a multilayer film reflective mirror with a low internal stress in which the drop in reflectivity is suppressed can be obtained. The third invention that is used to achieve the object described above is an X-ray exposure apparatus which comprises an X-ray light source that generates X-rays, an illumination optical system that conducts X-rays from this X-ray light source to a mask, and a projection optical system that conducts the X-rays from the mask to a photosensitive substrate, and which transfers the pattern of the mask to the photosensitive substrate, wherein at least one part selected from the illumination optical system, mask and projection optical system has one of the multilayer film reflective mirrors described above. In the present invention, the internal stress can be reduced while suppressing a drop in the reflectivity of the multilayer film reflective mirror; accordingly, a deterioration in the optical characteristics can be prevented, so that a high-performance X-ray exposure apparatus can be obtained. As a result of investigating the problems encountered in the prior art described above, the present inventor obtained the following findings: In a multilayer film (e.g., an Mo/Si multilayer film) formed by alternately laminating layers consisting of a substance in which the difference between the refractive index of this substance in the X-ray region and the refractive index of a vacuum is large (first substance, e.g., Mo) and layers consisting of a substance in which this difference is small (second substance, e.g., Si), the Mo layers tend to undergo micro-crystallization, and, depending on the film formation method used, micro-crystallization may become more conspicuous as the thickness of the Mo layers increases, so that the surface roughness of the Mo layers increases. For example, the surface roughness when the thickness of the Mo layers is approximately 2.5 nm (Γ=0.35) is approximately 0.25 nmRMS, the surface roughness when the thickness of the Mo layers is approximately 3.6 nm (Γ=0.5) is approximately 0.34 nmRMS, the surface roughness when the thickness of the Mo layers is approximately 4.3 nm (Γ=0.6) is approximately 0.49 nmRMS, and the surface roughness when the thickness of the Mo layers is approximately 5 nm (Γ=0.7) is approximately 0.61 nmRMS; thus, the surface roughness increases with an increase in the thickness of the Mo layers. Furthermore, when the surface roughness of the Mo layers increases as a result of micro-crystallization, the reflectivity of the multilayer film reflective mirror drops. For example, the reflectivity when the thickness of the Mo layers is approximately 2.5 nm (Γ=0.35) is approximately 70%, the reflectivity when the thickness of the Mo layers is approximately 3.6 nm (Γ=0.5) is approximately 65%, the reflectivity when the thickness of the Mo layers is approximately 4.3 nm (Γ=0.6) is approximately 55%, and the reflectivity when the thickness of the Mo layers is approximately 5 nm (Γ=0.7) is approximately 40%; thus, the reflectivity of the multilayer film reflective mirror drops as the thickness of the Mo layers increases. In conventional stress reduction techniques, a first multilayer film which has a tensile stress is installed in order to reduce the compressive stress of the second multilayer film which has a high reflectivity. The periodic length of this first multilayer film is substantially the same as the periodic length of the second multilayer film, and Γ is relatively large (e.g., Γ is approximately 0.7); accordingly, the thickness of the Mo layers in the first multilayer film is increased. For instance, referring to FIG. 5, in a conventional low-stress multilayer film reflective mirror 51, the thickness of the Mo layers 551 in the second multilayer film 55 is approximately 2.5 nm (7.2 nm×0.35), while the thickness of the Mo layers 571 in the first multilayer film 57 is approximately 5 nm (7.2 nm×0.7), so that the thickness of the Mo layers 571 is increased compared to the thickness of the Mo layers 551. Since micro-crystallization becomes conspicuous as the thickness of the Mo layers increases, the surface roughness caused by the micro-crystallization of the Mo layers 571 increases. When the surface roughness caused by the micro-crystallization of the Mo layers 571 thus increases, the reflectivity of the second multilayer film 55 that is formed on top of the first multilayer film 57 drops. As a result, the reflectivity of the multilayer film reflective mirror 51 drops. Accordingly, in the present invention, a method was adopted in which Γ is increased without greatly increasing the thickness of the Mo layers in the first multilayer film. Specifically, in the multilayer film reflective mirror of the present invention, the thickness of the Mo layers in the first multilayer film is substantially the same as the thickness of the Mo layers in the second multilayer film having a high reflectivity, or the thickness of the Mo layers in the first multilayer film is smaller than the thickness of the Mo layers in the second multilayer film. Furthermore, the thickness of the Si layers in the first multilayer film is smaller than the thickness of the Si layers in the second multilayer film. Working configurations of the present invention will be described below with reference to the figures. FIG. 1 is a sectional view which shows the structure of a multilayer film reflective mirror constituting a working configuration of the present invention in model form. In this multilayer film reflective mirror 61, a first multilayer film 67 is formed between a substrate 63 and a second multilayer film 65. The second multilayer film 65 is an Mo/Si multilayer film consisting of Mo layers 651 and Si layers 653; this multilayer film 65 has a periodic length of 7.2 nm, a Γ value of 0.35, and a laminated layer number of 50 layer pairs, thus making it possible to obtain a high X-ray reflectivity. Meanwhile, the first multilayer film 67 is an Mo/Si multilayer film consisting of Mo layers 671 and Si layers 673; this multilayer film 67 has a periodic length of 3.6 nm, a Γ value of 0.7 and a laminated layer number of 88 layer pairs. Furthermore, in order to simplify the description, in the figure, a portion of the multilayer film reflective mirror is made horizontal, and the number of laminated layers is abbreviated. In this multilayer film reflective mirror 61, the thickness of the Mo layers 651 in the second multilayer film 65 is approximately 2.5 nm (7.2 nm×0.35). Furthermore, the thickness of the Mo layers 671 in the first multilayer film 67 is approximately 2.5 nm (3.6 nm×0.7), so that this thickness is set equal to the thickness of the Mo layers 651 in the second multilayer film 65. Moreover, these thicknesses may also differ somewhat as long as this difference is in a range which is such that there is no effect on the reflectivity of the multilayer film reflective mirror. Furthermore, the thickness of the layers of the first substance (Mo layers) in the first multilayer film may also be made smaller than the thickness of the layers of the first substance (Mo layers) in the second multilayer film. The thickness of the layers of the first substance in the first multilayer film must be adjusted in accordance with the stress that is required in order to reduce the stress in the second multilayer film, the work that is required for multilayer film formation, and the like. If the thickness of the layers of the first substance in the first multilayer film is less than 50% of the thickness of the layers of the first substance in the second multilayer film, the formation of the multilayer film may become difficult, and it may become impossible to obtain the stress that is required in order to reduce the stress in the second multilayer film. Furthermore, the work required for forming the multilayer film is increased, e.g., the number of laminated layers in the multilayer film is greatly increased in order to obtain the necessary stress. On the other hand, if the thickness of the layers of the first substance in the first multilayer film exceeds 120% of the thickness of the layers of the first substance in the second multilayer film, the surface roughness caused by micro-crystallization exceeds the permissible value (approximately 0.3 nmRMS), so that there is an effect on the reflectivity of the multilayer film reflective mirror. Accordingly, it is desirable that the thickness of the layers of the first substance in the first multilayer film be 50% to 120% of the thickness of the layers of the first substance in the second multilayer film. Furthermore, if the thickness of the layers of the first substance in the first multilayer film is set at 75% to 115% of the thickness of the layers of the first substance in the second multilayer film, the multilayer film can be formed more easily, and the surface roughness can be more securely kept to the permissible value or less, so that the effect on the reflectivity can be reduced; accordingly, such a thickness is even more desirable. Here, the EUV reflectivity of the multilayer film reflective mirror 61 will be discussed. The reflectivity of an Mo/Si multilayer film increases with the number of laminated layers; when a certain fixed number of layers is exceeded, this reflectivity reaches saturation and becomes fixed. The number of laminated layers in the second multilayer film 65 is approximately 40 layer pairs to 50 layer pairs, which is a sufficient number of layers for the reflectivity to become saturated. Accordingly, even if the periodic length of the first multilayer film 67 located beneath the second multilayer film 65 is not set so that this film has a high reflectivity with respect to the EUV wavelength used, the EUV reflectivity of the multilayer film reflective mirror 61 is maintained “as is” at a high value. Furthermore, the thickness of the Mo layers 671 in the first multilayer film 67 is set equal to the thickness of the Mo layers 651 in the second multilayer film 65. Accordingly, there is no drop in reflectivity caused by an increase in the thickness of the Mo layers. Thus, in the multilayer film reflective mirror 61 of the present invention, such a drop in the EUV reflectivity is suppressed, so that the mirror has a high reflectivity. Next, the internal stress of the multilayer film reflective mirror 61 will be discussed. Even in the case of a multilayer film in which Γ is increased and the periodic length is shortened (e.g., to 3.6 nm) without varying the thickness of the Mo layers, as in the first multilayer film 67, it has been confirmed that multilayer films with a large Γ value have a tensile stress. Accordingly, by combining the second multilayer film 65 which has a compressive stress and the first multilayer film 67 which has a tensile stress, it is possible to reduce the internal stress of the multilayer film reflective mirror 61. In this case, the force that is applied to the substrate increases with an increase in the film thickness even at the same stress; therefore, it is advisable to consider not only the magnitude of the stress in the respective multilayer films, but also the “total stress” which is the product of the film thicknesses of the individual multilayer films and the magnitude of the stress. After the stress values for the individual multilayer films are measured, the number of laminated layers of the first multilayer film 67 may be appropriately selected so that the total stress of the second multilayer film 65 and the total stress of the first multilayer film 67 are balanced. For example, where d2 is the periodic length of the second multilayer film 65, N2 is the number of laminated layers, and S2 is the stress, the total stress T2 of the second multilayer film 65 is expressed as follows:T2=d2×N2×S2Similarly, where d1 is the periodic length of the first multilayer film 67, N1 is the number of laminated layers, and S1 is the stress, the total stress T1 of the first multilayer film 67 is expressed as follows:T1=d1×N1×S1Accordingly, by selecting the number of laminate layers N1 of the first multilayer film 67 so that T1+T2=0, it is possible to cancel the internal stress of the multilayer film reflective mirror 61. However, there are cases in which the stress also varies in accordance with the number of laminated layers, and cases in which the stress is not completely canceled by the number of laminated layers selected as described above. In such cases, it is necessary to adjust the number of laminated layers of the first multilayer film 67 in accordance with the residual stress. Furthermore, as was described above, since the thickness of the Mo layers 671 of the first multilayer film 67 is not varied, there is no drop in reflectivity caused by an increase in the thickness of the Mo layers. Thus, a multilayer film reflective mirror with a low internal stress in which the drop in reflectivity is suppressed can be obtained by combining a second multilayer film 65 and a first multilayer film 67 in which the thickness of the Mo layers is substantially the same, but the Γ value is different. Embodiment 1 In Embodiment 1 of the present invention, a second multilayer film 65 and a first multilayer film 67 were formed using a low-voltage discharge cathode type rotary magnet cathode sputtering apparatus (a type of direct-current magnetron sputtering apparatus). The film formation conditions were as follows: namely, xenon (Xe) was used as the sputtering gas, the Xe gas flow was set at 3 sccm (0.08 Pa), and the cathode power was set at 200 W in the case of Mo, and 400 W in the case of Si. The second multilayer film 65 was an Mo/Si multilayer film in which the periodic length d2 was set at 7.2 nm, Γ was set at 0.35, and the number of laminated layers N2 was set at 50 layer pairs in order to obtain a high X-ray reflectivity. When the EUV reflectivity of this second multilayer film 65 was measured, a value of approximately 69% was obtained. Furthermore, the stress S2 of the second multilayer film 65 was a compressive stress of −350 MPa. In order to reduce the compressive stress of the second multilayer film 65, an Mo/Si multilayer film with a periodic length d1 of 3.6 nm and a Γ value of 0.7 was selected as the first multilayer film 67 having a tensile stress. The thickness of the Mo layers 671 in this first multilayer film 67 was the same as the thickness of the Mo layers 651 in the second multilayer film 65. When the stress S1 of the first multilayer film 67 was measured, a tensile stress value of +400 MPa was obtained. Here, the method used to determine the number of laminated layers N1 in the first multilayer film 67 will be described. The number of laminated layers N1 in the first multilayer film 67 was determined as shown below on the basis of the stress values of the second multilayer film 65 and first multilayer film 67. The total stress T2 of the second multilayer film 65 in which the number of laminated layers is 50 layer pairs is as follows: T2 = d2 × N2 × S2 = ( 7.2 ⁢ ⁢ nm ) × ( 50 ) × ( - 350 ⁢ ⁢ MPa ) = - 126 ⁢ ⁢ N ⁢ / ⁢ m From the conditional formula T1+T2=0 used to reduce the internal stress of the multilayer film reflective mirror 61 and the formula T1=d1×N1×S1 for determining the total stress T1 of the first multilayer film 67, the number of laminated layers N1 of the first multilayer film 67 is as follows: N1 = ⁢ ( + 126 ⁢ ⁢ N ⁢ / ⁢ m ) / { ( 3.6 ⁢ ⁢ nm ) × ( + 400 ⁢ ⁢ MPa ) } = ⁢ 87.5 ≈ ⁢ 88 Furthermore, since the number of laminated layers N1 must be set as an integer, decimal fractions are rounded upward, so that N1=88 layer pairs. Accordingly, as is shown in FIG. 1, a first multilayer film 67 in which the number of laminated layers was 88 layer pairs was formed on the substrate 63, and a second multilayer film 65 in which the number of laminated layers was 50 layer pairs was formed on top of this first multilayer film 67. When the stress of this multilayer film reflective mirror 61 was measured, it was found that this stress had been reduced to 20 MPa or less. Furthermore, the stress of the multilayer film reflective mirror 61 can be reduced to a value close to zero by a further fine adjustment of the number of laminated layers N1 of the first multilayer film 67. Moreover, the surface roughness of the multilayer film reflective mirror 61 was approximately 0.26 nmRMS, and was thus kept below the permissible value (approximately 0.3 nmRMS). Furthermore, the reflectivity of the multilayer film reflective mirror 61 was 69%, so that almost no drop in reflectivity was recognized. For purposes of comparison, the internal stress and reflectivity in a case where a multilayer film reflective mirror was manufactured using a conventional stress reduction technique were also determined. This conventional multilayer film reflective mirror was, for example, the multilayer film reflective mirror 51 shown in FIG. 5. The second multilayer film 55 of the multilayer film reflective mirror 51 was similar to the second multilayer film 65 of the multilayer film reflective mirror 61 of the present invention. Accordingly, the total stress T4 of the second multilayer film 55 in which the number of laminated layers was 50 layer pairs was as follows: T4 = ( 7.2 ⁢ ⁢ nm ) × ( 50 ) × ( - 350 ⁢ ⁢ MPa ) = - 126 ⁢ ⁢ N ⁢ / ⁢ m Assuming that the periodic length d3 of the first multilayer film 57 was 7.2 nm, and that the stress S3 was a tensile stress of +600 MPa, the number of laminated layer N3 in the first multilayer film 57 required in order to reduce the internal stress of the multilayer film reflective mirror 51 was as follows: N3 = ⁢ ( + 126 ⁢ ⁢ N ⁢ / ⁢ m ) / { ( 7.2 ⁢ ⁢ nm ) × ( + 600 ⁢ ⁢ MPa ) } ≈ ⁢ 30 Accordingly, as is shown in FIG. 5, a first multilayer film 57 in which the number of laminated layers was 30 layer pairs was formed on the substrate 53, and a second multilayer film 55 in which the number of laminated layers was 50 layer pairs was formed on top of this first multilayer film 57. When the stress of this multilayer film reflective mirror 51 was measured, it was found that the stress had been reduced to 20 MPa or less. However, the reflectivity of the multilayer film reflective mirror 51 dropped to 63%. It appears that the reason for this was that since the thickness of the Mo layers 571 in the first multilayer film 57 of the multilayer film reflective mirror 51 was large, i.e., approximately 5 nm (7.2 nm×0.7), the surface roughness increased as a result of micro-crystallization of the Mo layers. Accordingly, when the surface roughness of the multilayer film reflective mirror 51 was measured, the value obtained was approximately 0.39 nmRMS, which exceeded the permissible value (approximately 0.3 nmRMS) for the surface roughness. Thus, it was found that the multilayer film reflective mirror 51 in FIG. 5 using a conventional stress reduction technique showed a drop in reflectivity, while the multilayer film reflective mirror 61 of the present invention shown in FIG. 1 allowed a reduction in stress to be achieved while maintaining a high reflectivity. Embodiment 2 In Embodiment 2 of the present invention, a second multilayer film 65 and a first multilayer film 67 were formed using a low-voltage discharge cathode type rotary magnet cathode sputtering apparatus (a type of direct-current magnetron sputtering apparatus). The film formation conditions were as follows: namely, xenon (Xe) was used as the sputtering gas, the Xe gas flow was set at 3 sccm (0.08 Pa), and the cathode power was set at 200 W in the case of Mo, and 400 W in the case of Si. The second multilayer film 65 was an Mo/Si multilayer film in which the periodic length d2 was set at 7.2 nm, Γ was set at 0.35, and the number of laminated layers N2 was set at 50 layer pairs in order to obtain a high X-ray reflectivity. When the EUV reflectivity of this second multilayer film 65 was measured, a value of approximately 69% was obtained. Furthermore, the stress S2 of the second multilayer film 65 was a compressive stress of −350 MPa. In order to reduce the compressive stress of the second multilayer film 65, an Mo/Si multilayer film with a periodic length d1 of 2.9 nm and a Γ value of 0.75 was selected as the first multilayer film 67 having a tensile stress. The thickness of the Mo layers 671 in this first multilayer film 67 was smaller than the thickness of the Mo layers 651 in the second multilayer film 65. When the stress S1 of the first multilayer film 67 was measured, a tensile stress value of +300 MPa was obtained. Here, the method used to determine the number of laminated layers N1 in the first multilayer film 67 will be described. The number of laminated layers N1 in the first multilayer film 67 was determined as follows on the basis of the stress values of the second multilayer film 65 and first multilayer film 67. The total stress T2 of the second multilayer film 65 in which the number of laminated layers is 50 layer pairs is as follows: T2 = d2 × N2 × S2 = ( 7.2 ⁢ ⁢ nm ) × ( 50 ) × ( - 350 ⁢ ⁢ MPa ) = - 126 ⁢ ⁢ N ⁢ / ⁢ m From the conditional formula T1+T2=0 used to reduce the internal stress of the multilayer film reflective mirror 61 and the formula T1=d1×N1×S1 for determining the total stress T1 of the first multilayer film 67, the number of laminated layers N1 of the first multilayer film 67 is as follows: N1 = ⁢ ( + 126 ⁢ ⁢ N ⁢ / ⁢ m ) / { ( 2.9 ⁢ ⁢ nm ) × ( + 300 ⁢ ⁢ MPa ) } ≈ ⁢ 145 Accordingly, as is shown in FIG. 1, a first multilayer film 67 in which the number of laminated layers was 145 layer pairs was formed on the substrate 63, and a second multilayer film 65 in which the number of laminated layers was 50 layer pairs was formed on top of this first multilayer film 67. When the stress of this multilayer film reflective mirror 61 was measured, it was found that this stress had been reduced to 20 MPa or less. Furthermore, the stress of the multilayer film reflective mirror 61 can be reduced to a value close to zero by a further fine adjustment of the number of laminated layers N1 of the first multilayer film 67. Moreover, the surface roughness of the multilayer film reflective mirror 61 was approximately 0.26 nmRMS, and was thus kept below the permissible value (approximately 0.3 nmRMS). Furthermore, the reflectivity of the multilayer film reflective mirror 61 was 69%, so that almost no drop in reflectivity was recognized. Embodiment 3 In Embodiment 3 of the present invention, a second multilayer film 65 and a first multilayer film 67 were formed using a low-voltage discharge cathode type rotary magnet cathode sputtering apparatus (a type of direct-current magnetron sputtering apparatus). The film formation conditions were as follows: namely, xenon (Xe) was used as the sputtering gas, the Xe gas flow was set at 3 sccm (0.08 Pa), and the cathode power was set at 200 W in the case of Mo, and 400 W in the case of Si. The second multilayer film 65 was an Mo/Si multilayer film in which the periodic length d2 was set at 7.2 nm, Γ was set at 0.35, and the number of laminated layers N2 was set at 50 layer pairs in order to obtain a high X-ray reflectivity When the EUV reflectivity of this second multilayer film 65 was measured, a value of approximately 69% was obtained. Furthermore, the stress S2 of the second multilayer film 65 was a compressive stress of −350 MPa. In order to reduce the compressive stress of the second multilayer film 65, an Mo/Si multilayer film with a periodic length d1 of 4.6 nm and a Γ value of 0.65 was selected as the first multilayer film 67 having a tensile stress. When the stress S1 of the first multilayer film 67 was measured, a tensile stress value of +300 MPa was obtained. Here, the method used to determine the number of laminated layers N1 in the first multilayer film 67 will be described. The number of laminated layers N1 in the first multilayer film 67 was determined as follows on the basis of the stress values of the second multilayer film 65 and first multilayer film 67. The total stress T2 of the second multilayer film 65 in which the number of laminated layers is 50 layer pairs is as follows: T2 = d2 × N2 × S2 = ( 7.2 ⁢ ⁢ nm ) × ( 50 ) × ( - 350 ⁢ ⁢ MPa ) = - 126 ⁢ ⁢ N ⁢ / ⁢ m From the conditional formula T1+T2=0 used to reduce the internal stress of the multilayer film reflective mirror 61 and the formula T1=d1×N1×S1 for determining the total stress T1 of the first multilayer film 67, the number of laminated layers N1 of the first multilayer film 67 is as follows: N1 = ⁢ ( + 126 ⁢ ⁢ N ⁢ / ⁢ m ) / { ( 4.6 ⁢ ⁢ nm ) × ( + 300 ⁢ ⁢ MPa ) } ≈ ⁢ 92 Accordingly, as is shown in FIG. 1, a first multilayer film 67 in which the number of laminated layers was 92 layer pairs was formed on the substrate 63, and a second multilayer film 65 in which the number of laminated layers was 50 layer pairs was formed on top of this first multilayer film 67. When the stress of this multilayer film reflective mirror 61 was measured, it was found that this stress had been reduced to 20 MPa or less. Furthermore, the stress of the multilayer film reflective mirror 61 can be reduced to a value close to zero by a further fine adjustment of the number of laminated layers N1 of the first multilayer film 67. Moreover, the surface roughness of the multilayer film reflective mirror 61 was approximately 0.29 nmRMS, and was thus kept below the permissible value (approximately 0.3 nmRMS). Furthermore, the reflectivity of the multilayer film reflective mirror 61 was 69%, so that almost no drop in reflectivity was recognized. Embodiment 4 In Embodiment 4 of the present invention, a second multilayer film 65 and a first multilayer film 67 were formed using a low-voltage discharge cathode type rotary magnet cathode sputtering apparatus (a type of direct-current magnetron sputtering apparatus). The film formation conditions were as follows: namely, xenon (Xe) was used as the sputtering gas, the Xe gas flow was set at 3 sccm (0.08 Pa), and the cathode power was set at 200 W in the case of Mo, and 400 W in the case of Si. The second multilayer film 65 was an Mo/Si multilayer film in which the periodic length d2 was set at 7.2 nm, Γ was set at 0.35, and the number of laminated layers N2 was set at 45 layer pairs in order to obtain a high X-ray reflectivity. When the EUV reflectivity of this second multilayer film 65 was measured, a value of approximately 69% was obtained. Furthermore, the stress S2 of the second multilayer film 65 was a compressive stress of −350 MPa. In order to reduce the compressive stress of the second multilayer film 65, an Mo/Si multilayer film with a periodic length d1 of 3.3 nm and a Γ value of 0.7 was selected as the first multilayer film 67 having a tensile stress. The thickness of the Mo layers 671 in this first multilayer film 67 is substantially the same as the thickness of the Mo layers 651 in the second multilayer film 65. When the stress S1 of the first multilayer film 67 was measured, a tensile stress value of +408 MPa was obtained. Here, the method used to determine the number of laminated layers N1 in the first multilayer film 67 will be described. The number of laminated layers N1 in the first multilayer film 67 was determined as follows on the basis of the stress values of the second multilayer film 65 and first multilayer film 67. The total stress T2 of the second multilayer film 65 in which the number of laminated layers is 45 layer pairs is as follows: T2 = d2 × N2 × S2 = ( 7.2 ⁢ ⁢ nm ) × ( 45 ) × ( - 350 ⁢ ⁢ MPa ) = - 113 ⁢ ⁢ N ⁢ / ⁢ m From the conditional formula T1+T2=0 used to reduce the internal stress of the multilayer film reflective mirror 61 and the formula T1=d1×N1×S1 for determining the total stress T1 of the first multilayer film 67, the number of laminated layers N1 of the first multilayer film 67 is as follows: N1 = ⁢ ( + 113 ⁢ ⁢ N ⁢ / ⁢ m ) / { ( 3.3 ⁢ ⁢ nm ) × ( + 408 ⁢ ⁢ MPa ) } ≈ ⁢ 84 Accordingly, as is shown in FIG. 1, a first multilayer film 67 in which the number of laminated layers was 84 layer pairs was formed on the substrate 63, and a second multilayer film 65 in which the number of laminated layers was 45 layer pairs was formed on top of this first multilayer film 67. When the stress of this multilayer film reflective mirror 61 was measured, it was found that this stress had been reduced to 20 MPa or less. Furthermore, the stress of the multilayer film reflective mirror 61 can be reduced to a value close to zero by a further fine adjustment of the number of laminated layers N1 of the first multilayer film 67. For example, when the number of laminated layers N1 of the first multilayer film 67 was set at 130 layer pairs, the stress of the multilayer film reflective mirror 61 was successfully reduced to −6 MPa. Moreover, the surface roughness of the multilayer film reflective mirror 61 was approximately 0.26 nmRMS, and was thus kept below the permissible value (approximately 0.3 nmRMS). Furthermore, the reflectivity of the multilayer film reflective mirror 61 was 69%, so that almost no drop in reflectivity was recognized. Embodiment 5 Next, an outline of an X-ray exposure apparatus mounting the multilayer film reflective mirror shown in FIG. 1 will be described with reference to FIG. 2. FIG. 2 is a diagram showing the overall construction of the X-ray exposure apparatus of the present invention. This X-ray exposure apparatus is a projection exposure apparatus that performs an exposure operation by means of a step and scan system using light in the soft X-ray region in the vicinity of a wavelength of 13 nm (hereafter referred to as EUV light) as illuminating light used for exposure. A laser light source 3 is disposed in the furthest upstream part of the X-ray exposure apparatus 1. The laser light source 3 has the function of supplying laser light with a wavelength in the infrared region to visible region; for example, a YAG laser, excimer laser, or the like using semiconductor laser excitation is used. The laser light emitted from the laser light source 3 is focused by a focusing optical system 5, and reaches a laser plasma light source 7 disposed in the lower part. The laser plasma light source 7 can generate X-rays in the vicinity of a wavelength of 13 nm with good efficiency. A nozzle (not shown in the figure) is disposed in the laser plasma light source 7; this nozzle jets xenon gas. The jetting xenon gas is subjected to high-intensity laser light in the laser plasma light source 7. The xenon gas is elevated to a high temperature by the energy of the high-intensity laser light, and is excited to a plasma state; then, when the gas makes a transition to a low potential state, the gas emits EUV light. Since EUV light has a low transmissivity with respect to air, the light path is covered by a chamber (vacuum chamber) 9 so that the outside air is blocked. Furthermore, since debris is generated from the nozzle that emits the xenon gas, it is necessary to dispose the chamber 9 separately from other chambers. A rotary parabolic reflective mirror 11 coated with an Mo/Si multilayer film is disposed in the upper part of the laser plasma light source 7. The X-rays radiated from the laser plasma light source 7 are incident on the parabolic reflective mirror 11, and only X-rays in the vicinity of a wavelength of 13 nm are reflected in parallel toward the bottom of the exposure apparatus 1. A visible light cutting X-ray transmitting filter 13 consisting of beryllium (Be) with a thickness of 0.15 μm is disposed beneath the rotary parabolic reflective mirror 11. Of the X-rays that are reflected by the parabolic reflective mirror 11, only the desired X-rays at a wavelength of 13 nm pass through the transmitting filter 13. The area in the vicinity of the transmitting filter 13 is covered by a chamber 15, so that the outside air is blocked. An exposure chamber 33 is disposed beneath the transmitting filter 13. An illumination optical system 17 is disposed beneath the transmitting filter 13 inside the exposure chamber 33. The illumination optical system 17 is constructed from the reflective mirrors of a condenser system, the reflective mirrors of a fly-eye optical system, or the like; this optical system 17 shapes the X-rays that are input from the transmitting filter 13 into a circular arc shape, and provides illumination that is oriented toward the left side of the figure. An X-ray reflective mirror 19 is disposed on the left side of the illumination optical system 17 in the figure. This X-ray reflective mirror 19 is a circular rotary parabolic mirror in which the reflective surface 19a on the right side (with respect to the figure) has a concave shape; this mirror is held in a vertical position by a holding member. The X-ray reflective mirror 19 consists of a quartz substrate in which a reflective surface 19a is worked with a high degree of precision. A multilayer film of Mo and Si which has a high reflectivity for X-rays at a wavelength of 13 nm is formed on the reflective surface 19a. Furthermore, in cases where X-rays at a wavelength of 10 to 15 nm are used, a multilayer film combining a substance such as ruthenium (Ru) or rhodium (Rh), and a substance such as Si, Be or carbon tetraboride (B4C), may also be used. A light path bending reflective mirror 21 is disposed at an inclination to the right of the X-ray reflective mirror 19 (with respect to the figure). A reflective type mask 23 is horizontally disposed above the light path bending reflective mirror 21 so that the reflective surface of this mask 23 is at the bottom. The X-rays emitted from the illumination optical system 17 are reflected and focused by the X-ray reflective mirror 19, and then reach the reflective surface of the reflective type mask 23 via the light path bending reflective mirror 21. A reflective film consisting of a multilayer film is also formed on the reflective surface of the reflective type mask 23. A mask pattern corresponding to the pattern that is transferred to the wafer 29 is formed in this reflective film. The reflective type mask 23 is fastened to a mask stage 25 shown above the mask 23 in the figure. The mask stage 25 can move at least in the Y direction, so that X-rays reflected by the light path bending reflective mirror 21 successively illuminate the mask 23. A projection optical system 27 and the wafer 29 are successively disposed in the lower part of the reflective type mask 23. The projection optical system 27 consists of a plurality of reflective mirrors or the like; this optical system 27 reduces the pattern on the reflective type mask 23 to a specified reduction magnification (e.g., ¼), and focuses this pattern on the wafer 29. The wafer 29 is fastened by suction chucking or the like to a wafer stage 31 that can move in the X, Y and Z directions. A preliminary evacuation chamber 37 (load and lock chamber) is disposed to the exposure chamber 33 via a gate valve 35. A vacuum pump 39 is connected to the preliminary evacuation chamber 37, and the preliminary evacuation chamber 37 is evacuated to a vacuum by the operation of the vacuum pump 39. When an exposure operation is performed, the reflective surface of the reflective type mask 23 is illuminated with EUV light by the illumination optical system 17. In this case, the reflective projection optical system 27 is caused to perform a relative synchronized scan of the reflective type mask 23 and wafer 29 at a specified speed ratio determined by the reduction magnification of the projection optical system. As a result, the entire circuit pattern of the reflective type mask 23 is transferred by a step and scan system to each of a plurality of shot regions on the wafer 29. Furthermore, the chips of the wafer 29 are, for example, 25×25 mm square, and an IC pattern of 0.07 μm L/S can be exposed on the resist. By using the multilayer film reflective mirror 61 shown in FIG. 1 as the reflective mirrors in the X-ray exposure apparatus 1, it is possible to provide an X-ray exposure apparatus which has a high reflectivity without causing any deterioration in the optical performance, and to suppress a drop in the throughput.
050826026	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS According to this invention, a spent reprocessing extraction solvent can be regenerated without producing a precipitate by selectively removing cationic impurities by an aqueous acidic cleaning agent via a hydrophobic porous membrane and by selectively removing anionic impurities by an alkaline aqueous cleaning agent via a hydrophobic porous membrane at the same time. As an extraction solvent for fuel-reprocessing in nuclear power plants, tributyl phosphate (hereinafter referred to as "TBP") diluted with n-dodecane is usually used. Both TBP and n-dodecane are decomposed by an influence of radiation or nitric acid to mostly produce acids. Among decomposed products of TBP, the produced amount of dibutyl phosphate (hereinafter referred to as "DBP") is large, which results in giving a great influence on the extraction behavior. Further, metal ions such as zirconium ions, etc. show a tendency to remain in the solvent when DPB is present. Since these metal ions are radioactive, deterioration of the extraction solvent further proceeds. In the regeneration step of the solvent, it is necessary to remove both anionic impurities such as DBP, etc. and cationic impurities such as zirconium ions, etc. from the solvent. In order to remove DBP, an aqueous sodium carbonate solution is usually used. But since metal ions easily transfer to an aqueous phase in the absence of DBP, they enter into the aqueous sodium carbonate solution. But since the aqueous sodium carbonate solution is strongly alkaline, the metal ions are easily precipitated by forming metal hydroxides. In order to prevent the precipitation, the solvent (an organic phase) is separated from the aqueous phase via a hydrophobic porous membrane according to this invention. By employing such a process as removing impurities through the membrane, it becomes possible to separate and remove impurities by contacting two or more aqueous phases with the solvent. In this invention, a hydrophobic porous membrane is used between the solvent containing cationic and anionic impurities (or ions) and aqueous solution phases for removing these ions. The properties of hydrophobic are necessary for only passing ions through micro holes of the membrane. As the hydrophobic porous membrane, there can be used those made from synthetic or natural, organic or inorganic high polymer materials, for example, polyamides, polyacrylonitriles, polyethylenes, polyesters, polypropylenes, polysulfones, polyvinyl alcohols, polyvinyl chloride, glass, etc. Among these membranes, the use of hydrophobic hollow fibers is preferable in order to enlarge the contact area between the spent solvent and two or more aqueous phases and to make the removing efficiency the same as that obtained by the mixer-settler method. FIG. 1 shows schematically a structure of hollow fiber, wherein A denotes a porous membrane, B a hollow portion, and C pores. Hollow fibers are slightly different in properties dependent on starting materials and production methods but usually have an inner diameter of 250 to 350 .mu.m and a thickness of 50 to 150 .mu.m. The porous membrane A has a number of pores C penetrating from the inner side to the outer side as shown in FIG. 1 (b). The open area ratio of 50% or more can easily be obtained. Since the outer diameter of hollow fiber is 1 mm or less, a very large surface area can be obtained by using a bundle of hollow fibers. Since the pore size of hollow fiber is as small as about 0.05 to 0.5 .mu.m, water passing through the hollow portion of hollow fiber does not leak out to outside of the hollow fiber when the hollow fiber is made from a hydrophobic material. FIG. 2 schematically shows a state of passing an organic solvent phase outside of the membrane A and passing a water phase inside, that is, the hollow portion, of the membrane A. Since the material of the membrane is hydrophobic, the water phase does not enter into pores C. In contrast, the solvent enters into pores to produce interfaces D between the solvent phase and the water phase at the inside (B) of the membrane. Through the interface D, impurities in the solvent can be extracted with the water phase. When hollow fibers are used, it becomes possible to remove cations and anions contained in the solvent selectively and separately by contacting two or more aqueous phase ion removing agents with the solvent via the porous membranes of hollow fibers. FIG. 3 schematically shows a case when two kinds of ion removing agents are used. In one hollow fiber, an anion removing agent E is passed and DBP is mainly removed selectively. In another hollow fiber, a cation removing agent F is passed and metal ions of uranium (U), zirconium (Zr), etc. are mainly removed selectively. Since ions can be selectively removed from a spent (or deteriorated) solvent, this treatment can be conducted depending on properties of ions present in the spent solvent. Unlike the mixer-settler wherein one phase is dispersed in another phase, followed by separation of the two phases by allowing to stand, it is almost not necessary in this invention to consider the specific weights, viscosities, etc., which results in broadening the selection of range of cleaning agents (or ion removing agents). Further, since no moving portion is necessary in the apparatus used in this invention unlike the method of using a mixer, the maintenance of the apparatus is easy and trouble of the apparatus is rare. When a bundle of hollow fibers is immersed in a spent solvent and an alkaline aqueous solution which is an anion removing agent is passed through a part of the hollow fibers and an acidic aqueous solution which is a cation removing agent (or a complexing agent for forming a water-soluble metal complex) is passed through the rest of the hollow fibers, cations and anions in the spent solvent can be removed simultaneously without producing precipitates of metal hydroxides. Further, since no apparatus for standing the dispersed phase is necessary in this invention, there is hardly obtained an influence of insertion of solid substances from the outside. It is preferable to arrange hollow fibers passing the anion removing agent and those passing the cation removing agent as closely as possible. This is because when DBP is removed from the spent solvent, the metal ions easily transfer to the aqueous phase. When the hollow fibers are separated with too far distances, it becomes possible that the metal ions insert into the anion removing agent. If possible, it is preferable to place hollow fibers passing the cation removing agent neighboring to hollow fibers passing the anion removing agent. Examples of such arrangement of hollow fiber modules are shown in FIG. 4, wherein H denotes cation removing hollow fibers and G denotes anion removing hollow fibers. When deterioration of the spent solvent is remarkable, the arrangement of pattern I is preferable, but the removal efficiency of DBP cannot be taken too high. It is possible to use various arrangements of hollow fiber modules passing two kinds of ion removing agents separately in various proportions other than the pattern II shown in FIG. 4. Any arrangement can be employed depending on the state of operation of nuclear power plants. Since radionuclides are present in the spent solvent, the hollow fibers should have resistance to radiation to some extent. Further, since a small amount of nitric acid is dissolved in the spent solvent, the hollow fibers should have resistance to nitric acid as well as resistance to alkalis considering the use of an aqueous alkaline cleaning agent (anion removing agent). Considering the above-mentioned requirements, hollow fibers made from a synthetic resin such as polystyrene, polyvinyl chloride, polypropylene, polyethylene, polyacrylonitrile, or the like are preferable. Such materials can withstand the dose rate of 10.sup.4 Gy or more and can be used for one year sufficiently. If necessary, the hollow fibers may be exchanged at every routine inspection. The hollow fibers can be produced by a conventional process disclosed in, for example, Japanese Patent Examined Publication No. 47-8595. As the anion removing agent, there can be used an aqueous solution (about 0.1 to 2 N) of alkali metal hydroxides such as NaOH, KOH, etc. In such a case, it is not necessary to consider entrapment of metal ions in the spent solvent. Thus, it is not necessary to add carbonate ions and tartarate ions to the aqueous solution. It is also possible to use an organic alkaline aqueous solution of hydrazine and hydroxylamine as the anion removing agent. Hydrazine and hydroxylamine have a fear of explosion in the presence of nitric acid and are very unstable. But, when they are used by passing through hollow fibers, they can be used safely, since a danger of mixing them in other nitric acid system is very rare. Further, since these anion removing agents can be vaporized completely by pyrolysis, the use of them is excellent from the viewpoint of reducing the amount of solid waste. Further, when the selective separation is carried out by using hollow fibers, since no radioactive metal ions are contained in the resulting anion removing agent, the treatment therefor is easy. It is also possible to use other various organic bases such as ethanolamine, pyridine, aniline, etc. When ethanolamine is used as an anion removing agent, it can be recovered by distillation due to its chemical stability and reused again. Cations contained in the spent solvent include ions of zirconium, uranium, plutonium usually in 0.1 to 5 moles. As the cation removing agent, there can be used an acidic aqueous solution of a acid which can form complexes with high stability constants with zirconium, uranium, and the like cations. As the acid, there can be used an inorganic acid such as carbonic acid, etc.; an organic acid such as formic acid, oxalic acid, butyric acid, tartaric acid, etc., alone or as a mixture thereof. The use of carbonic acid is very effective for uranium cations. It is possible to store uranium with a high concentration in an aqueous phase in the presence of CO.sub.3.sup.2- or HCO.sub.3.sup.-. The use of an organic acid such as oxalic acid, tartaric acid, etc. is effective for removing zirconium. The use of an organic acid is preferable for reducing the volume of treated waste. When the metallic ions are removed from the cation removing agent after the use by an agglomeration precipitation or ion exchange method, the resulting cation removing agent can be used again. This invention is illustrated further in detail referring to FIG. 5, wherein a spent organic solvent is regenerated by using as ion removing agents an aqueous solution of sodium hydroxide and an aqueous solution of oxalic acid in a first step and an aqueous solution of sodium hydroxide and an aqueous solution of nitric acid in a second step. In an oxalic acid storing tank 1, 2N-oxalic acid aqueous solution 2 is stored, 1N-NaOH aqueous solution 4 is stored in a sodium hydroxide storing tank 3 and 0.1N-HNO.sub.3 aqueous solution 6 is stored in a dilute nitric acid storing tank 5. To individual storing tanks, pumps 7, 8 and 9 are connected, respectively, so as to pass individual aqueous solutions to a solvent regeneration apparatus. The solvent regeneration apparatus comprises a first step ion removing portion 10 and a second step ion removing portion 11. A spent organic solvent comprising 30% by volume of TBP and 70% by volume of n-dodecane containing impurities such as metal ions, DBP, etc. is fed to the first step ion removing portion through a solvent inlet 12. The 2N-oxalic acid aqueous solution 2 is fed to an oxalic acid forcing inlet 13 via the pump 7 and passed to hollow fibers 14 for passing oxalic acid. On the other hand, the sodium hydroxide aqueous solution used in the second step ion removing portion 11 is passed to hollow fibers 16 for passing sodium hydroxide via a sodium hydroxide forcing inlet 15. Both hollow fibers 14 and 16 are rearranged in a hollow fibers initial crossing portion 17 so as to have the pattern I shown in FIG. 4. In the first step ion removing portion 10, the metal ions such as U.sup.+2, Zr.sup.+2, etc. in the spent solvent flowed outside of the hollow fibers 14 and 16 are extracted with the oxalic acid solution passing through the hollow fibers 14, and anions such as those of DBP in the spent solvent are extracted with the sodium hydroxide solution passing through the hollow fibers 16. The hollow fibers are rearranged again at a hollow fiber final crossing portion 18 to collect hollow fibers of the same system. The oxalic acid solution is collected at an oxalic acid outlet 19 and the sodium hydroxide solution is collected at a sodium hydroxide outlet 20, respectively. Then, these ion-containing solutions are passed to a waste treating system so as to be treated depending on properties of these wastes. On the other hand, the spent solvent treated in the first step ion removing portion 10 is passed to the second step ion removing portion 11 via a solvent transporting pipe 21. To the second step ion removing portion 11, the sodium hydroxide aqueous solution is fed from the sodium hydroxide storing tank 3 via the pump 8 and entered into hollow fibers (not shown in the drawing) at a sodium hydroxide forcing inlet 22. Dilute nitric acid is passed to a nitric acid forcing inlet 23 from the dilute nitric acid storing tank 5 via the pump 9 and entered into hollow fibers (not shown in the drawing). The hollow fibers are rearranged at a hollow fiber initial crossing portion 24 so as to have the pattern II shown in FIG. 4. In the second step ion removing portion 11, precise separation of DBP by the sodium hydroxide aqueous solution, and precise separation of metal ions and extraction of the flowed ion removing agent by the dilute nitric acid are conducted by hollow fibers. At a hollow fiber final crossing portion 25, hollow fibers are collected separately depending on the systems and the sodium hydroxide aqueous solution is passed from a sodium hydroxide outlet 26 to the first step ion removing portion 10. The dilute nitric acid is passed to an acid recovering system via a dilute nitric acid outlet 27. The solvent from which the cations and anions have been removed is taken out of a regenerated solvent outlet 28 to use again as a regenerated solvent. According to this invention, DBP removing ratio is more than 98% and metal ion removing ratio is more than 95%. According to this example, the spent organic solvent can be regenerated by using an apparatus with a simple structure and with no moving portion. Further there is no fear of lowering the ion removing efficiency due to the formation of precipitates. Another example of this invention is explained referring to FIG. 6, wherein the ion removing agents are regenerated and reused so as to reduce the amounts of radioactive wastes produced. Hydrazine from a hydrazine regenerating device 29 is sent to a solvent regenerating device 31 via a pump 30. When the amount of hydrazine is insufficient due to a loss, hydrazine is supplemented from a hydrazine storing tank 32 by opening a valve 33. A 2N-tartaric acid solution from a tartaric acid solution regenerating device 34 is sent to the solvent regenerating device 31 via a pump 30'. When the amount of tartaric acid is insufficient due to a loss, tartaric acid is supplemented from a tartaric acid storing tank 35 by opening a valve 33'. Hydrazine is fed to a hollow fiber opening portion for hydrazine 36 in the solvent regenerating device 31. In this portion, only inlets of hollow fibers 37 for passing hydrazine are opened and thus hydrazine enters into the hollow fibers 37. On the other hand, the tartaric acid solution is fed to a hollow fiber opening portion for tartaric acid 38. In this portion, only inlets of hollow fibers 39 for passing the tartaric acid solution are opened and thus the tartaric acid solution enters into the hollow fibers 39. A spent organic solvent comprising 30% by volume of TBP and 70% by volume of n-dodecane containing as inpurities DBP and metal ions is passed to the solvent regenerating device 31 from a solvent inlet 42. In the solvent regenerating device 31, the hollow fibers 37 for passing hydrazine and the hollow fibers 39 for passing the tartaric acid solution are packed densely. Anions such as DBP, etc. are extacted with the hydrazine passing through the hollow fibers 37 and the metal ions are extracted with the tartaric acid solution passing through the hollow fibers 39. After such an operation, the solvent is flowed out from a solvent outlet 43 as a regenerated solvent. If necessary, the ion removing operation can be carried out by using one or more the same devices 31 as mentioned above. The hollow fibers passing hydrazine have openings at a hollow fiber opening portion 40 for hydrazine, from which the hydrazine extracting the anions is taken out and sent to the hydrazine regenerating device 29. Hydrazine (N.sub.2 H.sub.4 .multidot.H.sub.2 O) vaporizes at 118.5.degree. C. under 740 mm Hg, while DBP cannot vaporize under normal conditions. Therefore, the used hydrazine can be regenerated by distillation and reused repeatedly. On the other hand, the hollow fibers passing the tartaric acid solution have openings at a hollow fiber opening portion 41 for tartaric acid, from which the tartaric acid solution extracting the metal ions is taken out and sent to the tartaric acid solution regenerating device 34. The tartaric acid is regenerated in the device 34 and reused repeatedly. In the tartaric acid regenerating device 34, a cation exchange resin conventionally used is packed and the metal ions are removed by ion exchange. Since the amount of metal ions in the spent solvent is not so large in usual, the cation exchange resin can be used for a long time without exchanging. According to this example, there can be constructed an apparatus for regenerating a spent organic solvent with a simple structure and a small amount of wastes produced. Further, there is no fear of lowering the ion removing efficiency by producing precipitates. According to this invention, by using a substance which can selectively extract uranium ions and plutonium ions from a spent solvent dissolving metal ions into an aqueous phase, an ion removing operation can be carried out by passing an anion removing agent aqueous solution, an uranium and plutonium selectively removing agent aqueous solution and a cation removing agent aqueous solution through hollow fibers separately in contact with a spent solvent containing anions such as DBP and metal ions such as uranium ions, plutonium ions, zirconium ions, etc. via the porous membranes of hollow fibers. By this operation, uranium and plutonium which are nuclear fuel substances and other radio nuclides can be recovered separately. Further, by this, a loss of nuclear fuel substances in the reprocessing step and the like can be prevented. In addition, contamination of a waste treating system by plutorium which has a long half-value period can be prevented so as to make the waste treatment simple. According to this invention, since cations and anions can be separated selectively from a spent organic solvent, inhibition of separation of captured ions by precipitation of metal hydroxides under alkaline circumstances according to the prior art can be prevented.
	abstract	X-ray projection-exposure apparatus are disclosed that achieve a higher accuracy of pattern overlay than conventionally. An embodiment of such an apparatus includes an X-ray source, an illumination-optical system for irradiating an X-ray beam from the X-ray source onto a mask defining a pattern, a mask stage for holding the mask, a projection-optical system for directing the patterned beam from the mask to project an image of the pattern onto a resist-coated wafer, a wafer stage for holding the wafer, and a mark-position-detection system for detecting the position of a mark formed on the wafer. The center of an exposure-image field of the projection-optical system is located at a position displaced from the center axis of the projection-optical system, and the center axis of the mark-position-detection system is located laterally adjacent the exposure-image field with respect to the center axis of the projection-optical system.
	summary
	description	This application is based upon and claims the benefit of priority from prior Japanese Application No. 2009-167793, filed on Jul. 16, 2009, the entire contents of which are incorporated herein by reference. 1. Field of the Invention The present invention relates to a pattern observation method, more specifically to, a pattern observation method for observing semiconductor device patterns by irradiating them with charged particle beams. 2. Background Art In recent years, with a progress in fine patterning and multi-layer processing of semiconductor devices, a pattern of holes (contact holes, via holes, etc.) having a large aspect ratio have been used in many cases. To accommodate this trend, it is increasingly necessary to observe the processing state of a pattern bottom surface, for example, to confirm that an insulator on the bottom surface of a hole pattern is sufficiently removed or check if a width of the hole pattern bottom surface satisfies specifications. As such a fine structure observation method in the semiconductor process, an optical method and a method using a charged-particle beam are available. First, as one of the optical pattern observation methods, a method referred to as scatterometry is known (see, for example, Japanese Patent Application Laid-Open No. 2002-260994). By this method, a pattern to be observed is irradiated with light and then its shape is guessed from a spectrum of the reflected light. Since this method utilizes the diffraction of light, a target to be observed needs to be a regular pattern. Therefore, this method cannot observe irregular patterns such as those of system LSI device patterns. Even in a regular pattern, a local pattern such as an endmost pattern of it cannot be observed. Next, as one of the pattern observation methods by use of charged particle beams, a scanning electron microscope (SEM) is known which uses electron beams as a probe. For example, in the case of observing device pattern of a semiconductor device, the SEM irradiates the device pattern with an electron beam (primary electrons), thereby producing secondary electrons from a surface of the device pattern. Then, the SEM detects the secondary electrons with a detector and generates an observation image (hereinafter referred to as secondary electron image) from detected data of the secondary electrons. The secondary electron image can be used to observe a shape of the device pattern, thereby measuring a width of the device pattern. However, by an ordinary observation method by use of the SEM, it is difficult to clearly observe the bottom surface of patterns having a large aspect ratio by using the secondary electron image. The reasons will be described as follows. As one example, a line and space pattern (hereinafter referred to as L/S pattern) in which a line pattern and a space pattern are disposed periodically will be used as an observation target in the description. If the space pattern has a large width as compared to a height of the line pattern, that is, the aspect ratio is small, secondary electrons generated from a bottom surface of the space pattern (hereinafter referred to as space bottom surface), when it is irradiated with an electron beam, strike side walls of the line pattern little and are emitted to an outside of the L/S pattern, thus reaching the detector. Accordingly, the space bottom surface can be observed brightly. On the other hand, if the aspect ratio is large, the secondary electrons generated from the space bottom surface readily strike the side walls of the line pattern, thus reaching the detector less. Therefore, the space bottom surface cannot brightly be observed. Incidentally, a method is proposed for observing a space bottom surface brightly in a case where the space bottom surface is electrically continuous with a substrate kept to a ground potential and a line pattern is discontinuous with the substrate (see, for example, Japanese Patent Application Laid-Open No. 2002-270655). By this method, an entire L/S pattern serving as a sample is irradiated with an electron beam beforehand to positively charge the sample and then irradiated with the electron beam, thereby obtaining a secondary electron image. By this method, a space bottom can be observed brightly even if the aspect ratio is large. The reasons will be described as follows. If the entire L/S pattern is irradiated with an electron beam, the line pattern's top surface (hereinafter referred to as line top surface) is positively charged, but the space bottom surface is not charged because it is grounded. If the pattern is irradiated with the electron beam after such a charged state is established, the secondary electrons emitted from the space bottom surface are drawn out of the L/S pattern by the positive charge on the line top surface, thus reaching the detector. Thus, the space bottom surface can be observed brightly. According to a first aspect of the present invention, a pattern observation method for observing a pattern formed on an insulating film and having a concave pattern and a convex pattern is provided, the method including: irradiating both of the concave pattern and the convex pattern in the pattern with a first observing charged particle beam, to obtain a temporary image of the pattern having region information of a convex pattern region corresponding to the convex pattern and region information of a concave pattern region corresponding to the concave pattern; irradiating the convex pattern region in the pattern with an electric field forming charged particle beam having a first incident voltage and the concave pattern region in the pattern with an electric field forming charged particle beam having a second incident voltage different from the first incident voltage based on the region information of the convex pattern region and the region information of the concave pattern region respectively, thereby forming an electric field between a top surface of the convex pattern and a bottom surface of the concave pattern so that charged particles emitted from the bottom surface of the concave pattern when this bottom surface is irradiated with the charged particle beam may be drawn out of the pattern; and irradiating both of the concave pattern and the convex pattern in the pattern with a second observing charged particle beam after the electric field is formed, to obtain an image of the pattern having the information of the bottom surface of the concave pattern. According to a second aspect of the present invention, a pattern observation method for observing a line and space pattern formed on an insulating film and having a line pattern and a space pattern is provided, the method including: irradiating both of the line pattern and the space pattern in the line and space pattern with a first observing electron beam, to obtain a temporary image of the line and space pattern having region information of a line pattern region corresponding to the line pattern and region information of a space pattern region corresponding to the space pattern; irradiating the line pattern region with an electric field forming electron beam having a first incident voltage and the space pattern region with an electric field forming electron beam having a second incident voltage which gives a smaller secondary electron emission coefficient than that by the first incident voltage based on the region information of the line pattern region and the region information of the space pattern region respectively, thereby forming an electric field between a top surface of the line pattern and a bottom surface of the space pattern so that secondary electrons emitted from the bottom surface of the space pattern when this bottom surface is irradiated with the electron beam may be drawn out of the line and space pattern; and irradiating both of the line pattern and the space pattern in the line and space pattern with a second observing electron beam after the electric field is formed, to obtain an image of the line and space pattern having the information of the bottom surface of the space pattern. Before the description of embodiments of the present invention, a description will be given of the background to the present inventor developing the present invention. According to the aforesaid method described in Japanese Patent Application Laid-Open No. 2002-270655, even if a pattern has a large aspect ratio, its space bottom surface can be observed clearly. However, this method, which is effective in a case where the space bottom surface is electrically continuous to a substrate kept to a predetermined potential, is not effective in a case where the space bottom surface is floating in potential. As an example in which the space bottom surface is floating in potential, for example, an L/S pattern may be given which is all composed of an insulating film. In a process of manufacturing a nowadays semiconductor device having a multi-layer structure, in many cases, to form an upper-layer interconnection, an insulating film is deposited and then a trench is formed in this insulating film, thereby forming an L/S pattern not electrically continuous to the substrate. As another example in which the space bottom surface is floating in potential, a mask pattern in a so-called sidewall processing process may be given. The sidewall processing process refers to a semiconductor process capable of forming a pattern smaller than the minimum resolution of an exposure device. One example of this semiconductor process will be described as follows. First, a pattern is formed with lithography on an insulating film which serves as a mask material and then thinned by wet etching. On sidewalls of this thinned pattern, an insulating film (sidewall insulating film) is formed. Then, the thinned pattern that lie between the sidewall insulating films are removed. Thus, an L/S pattern is formed which has a pitch not smaller than the minimum resolution of the exposure device. The L/S pattern will be used as a mask pattern, so that it is extremely important that the thinned pattern between the sidewall insulating films should be removed sufficiently. Next, a description will be given, with reference to FIGS. 11 and 12, of the reasons why a pattern bottom surface cannot clearly be observed if it is floating in potential. FIG. 11(a) shows a plan view of an L/S pattern 11 which is all composed of an insulating film. FIG. 11(b) is a cross-sectional view taken along line A-A of FIG. 11(a). As shown in FIGS. 11(a) and 11(b), in the L/S pattern 11, a line pattern 12 and a space pattern 13 are disposed alternately. Observation regions of the L/S pattern 11 are irradiated with an electron beam under positive charging conditions, that is, its secondary electron emission coefficient exceeds one. As a result, as can be seen from FIG. 12(a), the line pattern 12 and the space pattern 13 are both charged positively. Since a line top surface 12a and a space bottom surface 13a are charged to the same extent, there is no difference in potential between the line top surface 12a and the space bottom surface 13a, thus producing no electric field. Then, under observation conditions (for example, under conditions where the secondary electron emission coefficient is one), the L/S pattern 11 is scanned with an electron beam. Since there is no electric field between the line top surface 12a and the space bottom surface 13a, as can be seen from FIG. 12(a), effects are not obtained of drawing out secondary electrons emitted from the space bottom surface 13a toward a vacuum side (upper side in FIG. 12(a)), so that the secondary electrons cannot efficiently reach a detector. As a result, as can be seen from FIG. 12(b), in a secondary electron image, a line pattern region 14 is bright but a space pattern region 15 is dark, similar to the case of ordinary observation. Thus, if the space bottom surface 13a is not grounded, it cannot be observed brightly even by the aforesaid comparison example method. Accordingly, conventionally, in order to observe the bottom surface of a concave pattern, which is combined with a convex pattern on an insulating film, such an approach has been employed as to cleave a wafer and check its cross-sectional surfaces. However, this method takes much time to carry out and, moreover, has a problem in that in-line inspection is impossible. Here, in-line inspection refers to taking a sample wafer from a semiconductor device manufacturing process, observing patterns on the wafer, and placing it back to a manufacturing line. As described above, conventionally, a strong desire has been present for a pattern observation method that can clearly observe the bottom surface of a concave pattern formed in an insulating film. The present invention has been developed based on the aforesaid unique technological recognition of the present inventor and, as described in the following embodiments, provides a pattern observation method that can clearly observe the bottom surface of a concave pattern combined with a convex pattern on an insulating film. The present pattern observation method may be such as follows more specifically. First, both of a concave pattern and a convex pattern which are formed on an insulating film are irradiated with an observing charged particle beam. Thus, a temporary image of the concave and convex patterns is obtained which has region information of a convex pattern region corresponding to the convex pattern and region information of a concave pattern region corresponding to the concave pattern. It is to be noted that the region information contains information about at least any one of a position, a shape, and a dimension (size) of the convex pattern region (or concave pattern region). Subsequently, the convex pattern region is irradiated with an electric field forming charged particle beam having a first incident voltage and the concave pattern region is irradiated with an electric field forming charged particle beam having a second incident voltage different from the first incident voltage based on the region information of the convex pattern region and the region information of the concave pattern region respectively. Thus, an electric field is formed between a top surface of the convex pattern and a bottom surface of the concave pattern. This electric field is utilized to draw charged particles emitted from the bottom surface of the concave pattern when this bottom surface is irradiated with the observing charged particle beam, to the outside of the pattern. After such an electric field is formed, both of the concave pattern and the convex pattern in the concave-convex pattern are irradiated with an observing charged particle beam. Thus, an image is obtained of the concave-and-convex pattern having the information of the bottom surface of the concave pattern. In such a manner, even in a case where the bottom surface of the concave pattern is floating in potential, an electric field formed between the top surface of the convex pattern and the bottom surface of the concave pattern can be utilized so that charged particles emitted from the bottom surface of the concave pattern may be drawn out of the concave-convex pattern efficiently. As a result, the bottom surface of the concave pattern can be observed clearly. A description will be given below of three embodiments according to the present invention with reference to the drawings. It is to be noted that identical reference numerals are given to similar components in the drawings, and detailed description on the similar components will not be repeated. Before the description of a pattern observation method according to a first embodiment of the present invention, a description will be given of a relationship between an incident voltage (incident energy) of primary electrons and a secondary electron emission coefficient, with reference to FIG. 1. The secondary electron emission coefficient refers to a ratio of the number of emitted secondary electrons as against the number of primary electrons made incident upon a sample (=number of secondary electrons/number of primary electrons). FIG. 1 shows incident voltage dependency of the secondary electron emission coefficient of a silicon oxide film (SiO2). As can be seen from FIG. 1, insulating films such as a silicon oxide film may come under positive charging conditions where the secondary electron emission coefficient is in excess of one and negative charging conditions where the secondary electron emission coefficient is less than one. As can be seen from FIG. 1, for the silicon oxide films, a case where the incident voltage for primary electrons is lower than 100 V and a case where it is higher than 2000 V are negative charging conditions, and a case where the incident voltage is higher than 100 V and lower than 2000 V is positive charging conditions. If an insulating film is irradiated with an electron beam under the negative charging conditions, secondary electrons emitted from the sample are outnumbered by incident primary electrons because the secondary electron emission coefficient is less than one. Therefore, a region irradiated with the electron beam is negatively charged. On the other hand, if the insulating film is irradiated with an electron beam under the positive charging conditions, secondary electrons emitted from the sample outnumber incident primary electrons because the secondary electron emission coefficient is larger than one. Therefore, a region irradiated with the electron beam is positively charged. It is to be noted that conditions where the secondary electron emission coefficient is just one will be used preferably as sample observation conditions, that is, observation conditions because the sample will not be charged. In the present embodiment, a sample to be observed is an L/S pattern 1 composed of an insulating film (silicon oxide film shown in FIG. 2(a)). FIG. 2(b) is a cross-sectional view taken along line A-A of FIG. 2(a). As can be seen from FIGS. 2(a) and 2(b), the L/S pattern 1 has a line pattern 2 and a space pattern 3 which are disposed alternately. A description will be given below of a pattern observation method according to the present embodiment along a flowchart shown in FIG. 3. (1) First, an electron beam's incident voltage that provides the positive charging conditions and an electron beam's incident voltage that provides the negative charging conditions are determined (step S11). The secondary electron emission coefficient changes greatly with the material of a sample. Therefore, it is necessary to determine the incident voltages that provide the positive charging conditions and the negative charging conditions respectively in accordance with the material of the sample. For example, in the case of a silicon oxide film (SiO2), as can be seen from FIG. 1, the electron beam's incident voltage Vin that gives positive charging is 100<Vin<200[V], and the electron beam's incident voltage Vin that gives negative charging is Vin<100[V] or Vin>2000[V].(2) Next, the L/S pattern 1 is irradiated with the electron beam, to obtain a secondary electron image (temporary image) of an observation region (step S12). FIG. 4(a) shows a schematic diagram of the obtained secondary electron image. As can be seen from FIG. 4(a), a line pattern region 4 is bright but a space pattern region 5 is dark. This is because, as described above, secondary electrons emitted from a space bottom surface 3a of the L/S pattern 1 cannot efficiently be detected. In FIG. 4(a), a white strip-shaped region sandwiched between the line pattern region 4 and the space pattern region 5 indicates a sidewall 2b of the line pattern 2. As an incident angle of primary electrons with respect to the sample gets smaller, the number of the secondary electrons emitted from the sample increases, so that the sidewall 2b may be observed brighter than the line pattern and the space pattern. It is to be noted that preferably the value of the electron beam's incident voltage at the time of observing the sample surface in this process is selected so that the secondary electron emission coefficient may be one, in order to prevent the insulating film from being charged. For example, in a case where the insulating film is made of a silicon oxide, the incident voltage value should preferably be 100 V or 2000 V. (3) Next, the secondary electron image obtained at step S12 is used to calculate the line pattern region 4 corresponding to the line pattern 2 and the space pattern region 5 corresponding to the space pattern 3 (step S13). They may be calculated by a method that determines a region based on a difference in a gradation value of the L/S pattern, a method that extracts an outline of the pattern by utilizing image processing, etc. The thus calculated line pattern region 4 and space pattern region 5 are shown in FIG. 4(a). As can be seen from FIG. 4(a), the line pattern region 4 contains information about a position, a shape, and a dimension (line width) of the line pattern 2, and the space pattern region 5 contains information about a position, a shape, and a dimension (line width) of the space pattern 3.(4) Next, as can be seen from FIG. 4(a), the line pattern region 4 is irradiated with an electron beam under the positive charging conditions, and the space pattern region 5 is irradiated with an electron beam under the negative charging conditions (step S14). For example, the line pattern region 4 is irradiated with an electron beam having an incident voltage of 800 V, which is the positive charging conditions, and the space pattern region 5 is irradiated with an electron beam having an incident voltage of 2500 V or 50 V, which is the negative charging conditions. It is to be noted that these regions may be irradiated with the electron beam in an arbitrary order. A description will be given below of three methods about scanning by use of an electron beam. By the first scanning method, the L/S pattern 1 is scanned with an electron beam in its longitudinal direction. That is, in FIG. 4(a), a space pattern region 5 is scanned with the electron beam in the longitudinal direction (vertical direction in the figure) under the negative charging conditions and then, the incident voltage is switched to that for the positive charging conditions, to scan a line pattern region 4 adjacent to this space pattern region 5 with the electron beam in the longitudinal direction. This scanning is performed from the left to the right in the figure each time the incident voltage conditions are switched alternately. Thus, the entire observation region is irradiated with the electron beam. By the second scanning method, a plurality of space pattern regions 5 and a plurality of line pattern regions 4 are scanned as one group respectively. That is, first, only the plurality of space pattern regions 5 in the observation region may be scanned until all the space pattern regions 5 in the observation region are scanned under the negative charging conditions finally. Then, the incident voltage may be switched, to scan only the plurality of line pattern regions 4 in the observation region until all the line pattern regions 4 in the observation region are scanned under the positive charging conditions finally. Oppositely, first, only the line pattern regions 4 in the observation region are scanned until all the line pattern regions 4 in the observation region are scanned under the positive charging conditions finally. Then, only the space pattern regions 5 may be scanned under the negative charging conditions. This second scanning method has an advantage in that the incident voltage need not be switched quickly in accordance with the pattern. By the third scanning method, the L/S pattern 1 may be scanned with an electron beam in a direction perpendicular to it (horizontal direction in the FIG. 4(a)). In this case, the incident voltage is switched from that for the positive charging conditions to that for the negative charging conditions when the region irradiated with an electron beam is transferred from the line pattern region 4 to the space pattern region 5. On the other hand, when the irradiation region is transferred from the space pattern region 5 to the line pattern region 4, the incident voltage is switched from that for the negative charging conditions to that for the positive charging conditions. It is to be noted that scanning the observation region once by such a method is referred to as one frame. Generally, only one frame of scanning is not enough to obtain a desired charged state. That is, in order to obtain a desired charged state, typically, a plurality of frames of scanning is performed. In such a case of performing a plurality of frames of scanning, the charged state will be saturated gradually in accordance with a certain time constant. Therefore, as described in the subsequent processes, it is necessary to determine whether the charged state is saturated. (5) Next, gradation values of the line pattern region 4 and the space pattern region 5 are respectively calculated using the detected data of secondary electrons emitted from the L/S pattern 1 when it is irradiated with an electron beam at step S14 (step S15).(6) Next, a difference ΔL is calculated between the gradation value of the line pattern region 4 obtained at step S15 and that of the line pattern region 4 obtained at the previous frame. Similarly, a difference ΔS is calculated between the gradation value of the space pattern region 5 obtained at step S15 and that of the space pattern region 5 obtained at the previous frame (step S16). It is to be noted that when this process is performed for the first time (at the first frame), the gradation value obtained at step S15 is used as it is as the difference.(7) Next, the difference ΔL of the gradation value of the line pattern region 4 obtained at step S16 is compared to a preset threshold value XL. Similarly, the difference ΔS of the gradation value of the space pattern region 5 obtained at step S16 is compared to a preset threshold value XS. As a result, if the differences calculated at step S16 are smaller than the respective threshold values, that is, if ΔL<XL and ΔS<XS, the scanning ends to advance to step S18. Otherwise, the procedure returns to step S14 (step S17). The processing of steps S14 to S17 is thus repeated until the differences of the gradation values are lowered below the respective thresholds (predetermined values). In such a manner, a desired charged state is obtained. It is to be noted that there are possibilities that the differences of the gradation values may not be lowered below the respective threshold values even if the processing of steps S14 to S17 is repeated any number of times, depending on the material or structure of the sample. Therefore, an upper limit of the number of frames (the number of times of performing the processing of steps S14 to S16) is set beforehand so that even if the differences are not lowered below the threshold values, the procedure may advance to step S18 if the upper limit is reached. Such a charged state as shown in FIG. 4(b) is obtained by thus scanning the observation region as switching the incident voltage between the value for the positive charging conditions and that for the negative charging conditions in accordance with the patterns. FIG. 4(b) is a cross-sectional view of the L/S pattern, showing a charged state after this scanning is completed, that is, a charged state after the differences of the gradation values are lowered below the threshold values. As can be seen from FIG. 4(b), a line top surface 2a is charged positively and a space bottom surface 3a, negatively. By forming such a charged state, a local electric field is generated between the line top surface 2a and the space bottom surface 3a. (8) Next, the L/S pattern 1 is irradiated with an electron beam, to obtain a secondary electron image of the observation region (step S18). Similar to step S12, preferably the electron beam incident voltage is the value under the observation conditions (for example, 100 V or 2000 V). The local electric field formed by the charged state shown in FIG. 4(b) causes secondary electrons emitted from the space bottom surface 3a to be attracted by positive charge on the line top surface 2a, thus getting away to the side of vacuum (upper side in the figure). As a result, the secondary electrons emitted from the space bottom surface 3a can be detected efficiently by a detector. FIG. 4C is a schematic diagram of a secondary electron image obtained in this process. As can be seen from FIG. 4C, not only the line pattern region 4 but also the space pattern region 5 are observed brightly. In such a manner, in the present embodiment, the secondary electron image of an L/S pattern observed beforehand is used to calculate a space pattern region and a line pattern region. Then, the line pattern region is irradiated with an electron beam under the positive charging conditions and the space pattern region is irradiated with the electron beam under the negative charging conditions. Thus, a local electric field is formed between the top surface of a line pattern and the bottom surface of a space pattern. Then, an observation region is irradiated with the electron beam. Secondary electrons emitted from the space pattern's bottom surface are attracted by the local electric field, thus being taken out to the line pattern efficiently. As a result, by the pattern observation method according to the present embodiment, an efficiency is improved of detecting the secondary electrons, thus enabling clearly observing the bottom surface of the space pattern even if the L/S pattern is made of an insulating film. Next, a description will be given of a pattern observation method according to a second embodiment of the present invention. One of differences between the second embodiment and the first embodiment is a method of irradiation with an electron beam when generating a local electric field. In the first embodiment, the pattern has been irradiated with an electron beam whose focus position is set on the surface of a sample. In this case, to charge the sample sufficiently, a lot of scanning electron beams are required (for example, at least 400 scanning electron beams are necessary for one observation region). In contrast, in the present embodiment, an electron beam will be used whose focus position is shifted from the surface of the sample toward its inside so that the beam diameter of this electron beam on the top surface of the sample may be equal to a line width of the pattern. Moreover, besides the shifting of the focus position, an electron density of the electron beam is enhanced. Thus, the number of scanning electron beams required to sufficiently charge the sample will be decreased significantly. As a result, a time necessary in observation can be reduced greatly. A description will be given below of the pattern observation method according to the present embodiment along a flowchart shown in FIG. 5. Steps S21 to S23 are similar to steps S11 to S13 described with the first embodiment, and repetitive description on them will not be repeated, to start description from step S24. (1) A line pattern region 4 and a space pattern region 5 calculated at step S23 are used to calculate the line widths of a line pattern 2 and a space pattern 3 respectively (step S24). (2) Next, the focus position of an electron beam with which the line pattern region is to be irradiated is determined so that a beam diameter of this electron beam may be equal to the line width of the line pattern 2 calculated at step S24 on an irradiation surface (top surface of the line pattern 2). Similarly, the focus position of an electron beam with which the space pattern region is to be irradiated is determined so that a beam diameter of this electron beam may be equal to the line width of the space pattern 3 calculated at step S24 on an irradiation surface (bottom surface of the space pattern 3) (step S25). FIG. 6 shows a relationship between the schematic diagram of a secondary electron image obtained at S22 and the electron beam used in irradiation. In this example shown in FIG. 6, the space pattern region 5 has a larger line width than that of the line pattern region 4. Therefore, in the space pattern region 5, the focus position of the electron beam is greatly shifted from the irradiation surface toward the inside of a sample. On the other hand, in the line pattern region 4, the focus position is comparatively slightly shifted from the irradiation surface toward the inside of the sample. By thus adjusting the focus positions, the beam diameter of the electron beam matches each of the pattern line widths calculated at step S24. The following steps S26 to S30 are similar to steps S14 to S18 described with the first embodiment, and repetitive description on them will not be repeated. However, in irradiation at step S26, the electron beam used in it has a higher current than that in the first embodiment. It is done so in order to compensate for a decrease in number of electrons incident upon a unit area of the sample which decrease is caused by a shift in a focus position from the sample surface. As described already, typically, the number of scanning electron beams used for one observation region is at least 400. However, by the present embodiment, as described above, the focus position is adjusted so that the electron beam diameter may be equal to a line width of the pattern to be irradiated and, at the same time, the current of the electron beams is enhanced so that the number of electrons which are made incident upon the unit area may not decrease. Thus, only by scanning each of the line pattern region and the space pattern region once, one frame of irradiation with the electron beams can be completed to form a predetermined charged state. That is, in the case of an observation region including four space patterns and three line patterns as shown, for example, in FIG. 6, a predetermined charged state can be obtained only by scanning each of the patterns once. Therefore, one frame of scanning can be completed by scanning them with the electron beam only seven times in total. Thus, by the present embodiment, similar to the case of the first embodiment, even if the L/S pattern is made of an insulating film, the bottom surface of the space pattern can be observed clearly. Moreover, the present embodiment enables reducing the number of scanning electron beams significantly, thereby reducing an observation time of the sample greatly. So far, there have been described the two embodiments according to the present invention. In the description of the first and second embodiments, the line pattern (convex pattern) has been irradiated with an electron beam under the positive charging conditions and the space pattern (concave pattern) has been irradiated with an electron beam under the negative charging conditions. More generally, by the present invention, instead of whether it is under the positive charging conditions or the negative charging conditions, the line pattern and the space pattern may be irradiated with the electron beams under irradiation conditions having different secondary electron emission efficiencies. That is, the convex line pattern may be irradiated with the electron beam under the irradiation conditions having a secondary electron emission coefficient and the concave space pattern, under the irradiation conditions having a secondary electron emission coefficient η2 (η1>η2). Thus, an electric field to draw secondary electrons out of the L/S pattern to be observed is formed between the line pattern's top surface and the space pattern's bottom surface. Although the first and second embodiments have been described with reference to the example of using an L/S pattern as a pattern to be observed, the observation method according to the present invention can be applied also to any patterns other than the L/S pattern. That is, by the observation method according to the present invention, the bottom surface of a concave pattern on an insulating film having arbitrarily-shaped concave-convex pattern can be observed clearly. Next, a description will be given of a pattern observation method according to a third embodiment of the present invention. One of differences of the third embodiment from the first embodiment is that in contrast to the first embodiment, in which two irradiation regions (a line pattern region and a space pattern region) calculated from a secondary electron image of an observation region have been irradiated with electron beams under the different conditions, the present embodiment will calculate three irradiation regions from the secondary electron image of the observation region and irradiate them with the electron beams under the different conditions. By this method, even in a two-step hole pattern formed in an insulating film, that is, two-step hole pattern having a hole pattern (first hole pattern) and another hole pattern (second hole pattern) formed on the bottom surface of the first hole pattern, the bottom surface of each of those hole patterns can be observed clearly. Such a two-step hole pattern can be found in, for example, a dual damascene structure. FIG. 7 shows a structure of a two-step hole pattern 21 to be observed. FIG. 7(a) is a plan view of this hole pattern 21 and FIG. 7(b) is a cross-sectional view taken along line A-A of FIG. 7(a). As can be seen from FIGS. 7(a) and 7(b), in a top surface 21a of the two-step hole pattern 21, a first hole pattern 22 is formed, and in a bottom surface 22a of this first hole pattern 22, a second hole pattern 23 is formed. Next, a description will be given of the pattern observation method according to the present embodiment along a flowchart shown in FIG. 8. (1) First, an electron beam's incident voltage V1 that provides the first charging conditions, an electron beam's incident voltage V2 that provides the second charging conditions, and an electron beam's incident voltage V3 that provides the third charging conditions are determined (step S31). Assuming here that secondary electron emission efficiencies corresponding to the incident voltages V1, V2, and V3 to be η1, η2, and η3 respectively, the incident voltages V1, V2, and V3 are determined so that a relationship of η1>η2>η3 may be established. FIG. 9 shows one example of the thus determined relationship between V1, V2, and V3 and η1, η2 and η3.(2) Next, an entire observation region of the two-step hole pattern 21 is irradiated with an electron beam, to obtain a secondary electron image (temporary image) of the observation region (step S32). This secondary electron image has region information of a first region corresponding to the top surface 21a of the two-step hole pattern 21, region information of a second region corresponding to the bottom surface 22a of the first hole pattern 22, and region information of a third region corresponding to the bottom surface 23a of the second hole pattern 23.(3) Next, the secondary electron image obtained at step S32 is used to calculate a first region 24 corresponding to the top surface 21a of the two-step hole pattern 21, a second region 25 corresponding to the bottom surface 22a of the first hole pattern 22, and a third region 26 corresponding to the bottom surface 23a of the second hole pattern 23 (step S33). FIG. 10(a) shows the thus calculated first region 24, second region 25, and third region 26.(4) Next, the first region 24 is irradiated with an electron beam under the first charging conditions (incident voltage V1), the second region 25 is irradiated with an electron beam under the second charging conditions (incident voltage V2), and the third region 26 is irradiated with an electron beam under the third charging conditions (incident voltage V3) (step S34). It is to be noted that these regions may be irradiated with the electron beam in arbitrary order. The following steps S35 to S38 are similar to steps S15 to S18 described with the first embodiment, and repetitive description on them will not be repeated. FIG. 10(b) is a cross-sectional view of the two-step hole pattern 21, showing one example of the charged state of the two-step hole pattern 21 before the secondary electron image of the observation region is obtained at step S38. As can be seen from FIG. 10(b), the top surface 21a of the two-step hole pattern 21 is charged positively, the bottom surface 22a of the first hole pattern 22 is positively charged weaker than the top surface 21a, and the bottom surface 23a of the second hole pattern 23 is charged negatively. By forming such a charged state, local electric fields are respectively generated between the top surface 21a and the bottom surface 22a, between the bottom surface 22a and the bottom surface 23a, and between the top surface 21a and the bottom surface 23a. With this, when the secondary electron image is being obtained of the observation region of the two-step hole pattern 21 by irradiating it with the electron beam at step S38, secondary electrons emitted from the bottom surface 22a of the first hole pattern 22 are attracted by positive charge on the top surface 21a, thus getting drawn out to the side of vacuum (upper side in FIG. 10(b)). On the other hand, secondary electrons emitted from the bottom surface 23a of the second hole pattern 23 are attracted by positive charge on the top surface 21a and the bottom surface 22a, thus getting drawn out to the side of vacuum. As a result, the secondary electrons emitted from the bottom surfaces 22a and 23a can be detected efficiently. FIG. 10(c) is a schematic diagram of a secondary electron image of the observation region of the two-step hole pattern 21 obtained at step S38. As can be seen from FIG. 10(c), not only the top surface 21a of the two-step hole pattern 21 but also the bottom surfaces 22a and 23a of the respective first and second hole patterns 22 and 23 are observed brightly. It is to be noted that although the present embodiment has been described with reference to the two-step hole pattern, even in the case of a hole pattern having at least three steps, its respective bottom surfaces can be clearly observed by extending and applying the method according to the present embodiment. As thus described, by the pattern observation method according to the present embodiment, even in the case of a hole pattern having a shallow hole and a deep hole, bottom surfaces of those respective holes can be observed clearly. So far, there have been described the three embodiments according to the present invention. Although in the description of these embodiments, an electron beam has been used as a charged particle beam, any other charged particle beams (ion beam) than the electron beam may be used. Additional advantages and modifications will readily occur to those skilled in the art. Therefore, the invention in its broader aspects is not limited to the specific details and representative embodiments shown and described herein. Accordingly, various modifications may be made without departing from the spirit or scope of the general inventive concepts as defined by the appended claims and their equivalents.
039878600	abstract	A nuclear reactor core stabilizing arrangement wherein a plurality of actuators, disposed in a pattern laterally surrounding a group of elongated fuel assemblies, press against respective contiguous fuel assemblies on the periphery of the group to reduce the clearance between adjacent fuel assemblies thereby forming a more compacted, vibration resistant core structure.
	abstract	A device for turning a container about a turning axis is provided. The device includes a stationary holding member and two bearing members which are spaced apart along the turning axis on each side of a space for receiving a container, and which are capable of being used for bearing on the container while enabling the rotation of the container about the turning axis relative to the bearing members. The device includes a movable holding member that is mounted onto the stationary holding member so as to rotate about the turning axis, at least one bearing member being supported by the movable holding member while being mounted onto the movable holding member so as to rotate about the drive axis, the or each bearing member supported by the movable holding member not rotating about the turning axis relative to the stationary holding member when the movable holding member rotates about the turning axis relative to the stationary holding member.
	summary
	summary
	abstract	A method for designing masks adapted to the forming of integrated circuits, including the steps of: (a) forming a first test file including a set of configurations of integrated circuit elements; (b) forming a second test file comprising the elements of the first test file, less the elements corresponding to configurations forbidden by design rule manuals; (c) trans-forming the second test file by means of a set of logical operations implemented by computing means to obtain a mask file; (d) testing the mask file and, if the test is negative, modifying the design rule manuals; and (e) repeating steps (a) to (d) until the test of step (d) is positive.
	description	This application claims the benefit of U.S. Provisional Application No. 60/934,412 filled on Jun. 13, 2007, which is hereby incorporated by reference in this entity. 1. Field of the Invention The invention refers to structured materials used to trap and guide inside nuclear radiation or particles such as X- and gamma-rays, neutrons, charged particles, and their potential applications in nuclear power, imaging and shielding. 2. Description of the Related Art Channeling experiments proved that radiation may be trapped inside atomic lattices and steered in a similar manner with that the microwave and optical radiation is driven through wave-guides and, respectively, optical fiber. Charged particles and X-ray channeling has already reached the applications stage in accelerator and space technology. The new development based on nano-structures pushes the limits of channeling towards the high-energy radiation domain. The present solution of using complex nano-structures that can be electrically controlled opens the way to a new revolution in nuclear energy. A novel material that is able to trap and guide nuclear radiation in a controlled manner. The material is made of a plurality of controlled grown nano-structures, able to guide and gyrate the radiation along the structure at desired angle. The material may be built in hetero-structures inserting electric sensitive materials that make its channeling properties change with the applied voltage. A device made using such material that controls the radiation direction is possible of being used as control device in nuclear reactor replacing the existing control rods. FIG. 1—Shows three possibilities to make radiation shielding based on classical and non-classical approaches. The radiation sources 107 are supposed to emit radiation in space having a spherical symmetry, with the intensity at a distance r from the source given by the formula pictured in the figure above the source 107, where A is the source's activity for a particular radiation k. I0k is the particular radiation intensity of the radioactive source “k” placed in position “0”—the origin of the coordinate system, that hits a shielding element on the external surface placed at a distance “rk” from the radioactive source k. The ray 104 represents the path and the radiation Pointing vector that is hitting the shield and has the intensity on its surface I0k. The shield 103 has its thickness made of a plurality of elemental layers “i” 109 protecting a volume content 101 and has an associated coordinates system 102. At the contact with the surface 103 the incident radiation 104 is reflected 105 and refracted 106. The ratio Ir 105 to Io 104 gives the reflection coefficient, also called “albedo”, while the ratio It 106 to I0 104 gives the transmittsion coefficient. In current nuclear radiation calculations It is assumed equal with I0 because the reflection factor is smaller than 2% and is depending on incidence angle and energy. The classical absorption theory based on random interaction is following the exponential law having the absorption coefficient depending on material, density and radiation energy as showed by the formula pictured in the uppr left corner near the arrow labeled 106. The absorption length is defined as being the length where It becomes 1/e from Io or when the summation at the exponent of the absorption factors equals 1. There is important to understand that the absorbed radiation does not disappear, it is considered absorbed because it disappears from its original group “k” but the absorption location becomes a source reemitting the absorbed energies in other forms depending on the mechanism of interaction. The generic term of quantum scattering is covering the Thompson and Compton scattering as well the nuclear absorption and reemission also called non-elastic scattering, or resonant absorption and reemission of radiation. The quantum absorption, often called resonant absorption is based on the nuclear resonance mechanism where the radiation is exciting the inner energy levels of the nucleus or atom, saying that it excites the absorption element that further decays emitting the energy in various, other specific forms, becoming a secondary radiation source. In this moment the radiation associated particle say the gamma photon of energy Ek belonging to the specific group k disappears and other radiations k′ appears originating in that new place. The curve 110 shows the desired attenuation shape in the system of coordinates 102 of the shielded body 101 that relies on some path attenuation of the incident radiation 104, a high attenuation at the interaction with the shielding 103 that reduces the transmitted radiation 106 that further attenuates along the path inside the shielded volume 101. Drastic reduction of the transmitted radiation 106 at the incidence with the shielding material based on high reflection 105 due to a high reflection coefficient or “albedo” is impossible to be reached in usual cases for X, or gamma rays and neutrons due to their particularities of interaction with matter. The multi-layer shielding 103 made of individual successive material layers “i” 109 is showing the actual approach in radiation shielding, that has the disadvantage of low absorption factor that requires important material thickness and weight. FIG. 2—Shows the gamma radiation “degradation” (radiation energy gradual reduction) and absorption mechanism used in the actual best radiation shielding. The shielding 200 has a structure arbitrarily taken for exemplification purpose. The present shield is made of an inner layer of iron “Fe” 201, covered outside (shielded) by a layer of silver “Ag” 202 that is shielded at exterior by a layer of thorium “Th”203. As the chart 206 shows thorium having the biggest electronic density and mass density exhibits the biggest absorption coefficient for the supposed gamma radiation 204 generated by a radioactive source 205. If the gamma or X bremsstrahlung radiation energy is greater than 1.022 MeV the interaction process is dominated by the pair generation as shown in the plot 206 that has the vertical scale (ordinate) in relative units. The pair electron-positron shares the difference of energy (Erad−1022 keV) as kinetic energy. Both are stopping in the shielding matter by generating showers of knock-on electrons that generate X rays. When the positron energy becomes small enough it annihilates with a lattice electron reemitting the mass energy of 1022 KeV plus a share of the electron chemical bounding energy by two photons of a little bit more than 511 KeV. These photons released at near 180 deg. creates new gamma rays source localized somewhere in the Thorium 203 bulk. Another effect according to the plot 206 is the Compton effect. This effect is based on the collision between a photon with an electron that generates a recoiled electron and a lower energy photon. The recoiled electron is stopped in the lattice generating showers of knock-on electrons with associated X rays, while the lower energy photon in similar with a new source of radiation appeared somewhere inside the shielding, with some angular distribution, given by the Compton effect particularities. This is the mechanism that transforms a mono-energetic radiation into a plurality of radiations covering a large energy spectrum but with maximum energy lower than the energy of the initial radiation, conserving the energy. At lower energy materials start to exhibit specific absorption curves while at higher energies the mass density determines the absorption coefficient. The reasons for this “energy degrading” material combination is that Thorium 203 with high density has high stopping power is very effective to stop most of the initial beam 204 with high energy producing its own high energy X rays (i.e. 232Th-kα, kβ, L . . . , M . . . , etc), Compton and annihilation rays (511 keV). These Thorium secondary generated photns with lower energies are stopped in silver 202, which at its turn, emits more lower energies that are applied to Iron 201. The iron still emits 5.6 keV as K-alpha specific X ray higher energy and series of specific L, M lines. Aluminum—plastic linear may take out these lines leaving lines only lower than 1 keV and part of the entire attenuated spectrum. This means that each mono-energetic gamma ray is replaced after passing through the shielding with a continuum spectrum overlapped on a spectrum of characteristic lines of the elements it passed through, all at lower intensities. This kind of shielding leads to centimeters of material thickness and specific weight in tones. FIG. 3—Shows a procedure to enhance the attenuation coefficient by using a multi-layer resonant absorption attenuation arranged in an energy decreasing order as an embodiment of the present invention. As was mentioned in the explanations for FIG. 2, the attenuation coefficient of materials for photon radiation with energy over 500 keV is practically the same for about the same mass and only the density makes a difference in shielding thickness. To obtain a high attenuation requires excessive weight that impairs the applications. For neutrons, the shielding problem turns out to be more difficult, requiring several feet of absorbent materials around the radioactive source. Keeping in mind the Moessbauer effect and the most used elements in these experiments, there is possible that under the thorium outer layer in FIG. 2, 203 to introduce a plurality of Mossbauer elements layers, few mm thick each, with the resonant energy decreasing along the path like an “energy cascade”. These elements, generically called “Moessbauer elements” will enhance the overall absorption cross-section by the addition of the nuclear resonance that has as final effect an increase in scattering because a resonant photon absorption is followed by its reemission. In FIG. 3 the entire shield 300 is made from nuclear resonant layers arranged in the order of increasing the nuclear resonant energy towards the radioactive source 303 that produces the irradiation beam 302. The line 304 shows the assignment of the nuclear resonance energy in the plot 305 to a shield layer 301. The comb-looking nuclear resonances are coming to enhance the absorption cross-section of the multi-material shielding 300, finally reducing its thickness and weight. The theoretical predictions show a mass reduction of more than 50% from the initial shielding 200 in FIG. 2 to the new shielding 300 in FIG. 3, but this is not enough for many applications. In conclusion, by introducing centers of absorption and reemission of the radiation more than 50% of the incident radiation is backscattered and more than 80% or the radiation power may be taken by the shield. FIG. 4—Shows a main embodiment of the invention represented by the radiation “gyration” (term borrowed from microwave technology) schematic diagram obtained in bended nano-fibers acting like microwave-guides. The main idea developed in the patent is similar to the concept of radiation channeling in crystals. This concept is proven and in current use for charged particles and neutrons. More using elastic crystals there is possible to easy bend the beams of particles and neutrons similar to what happens to light in the optic-fibers. The development of nano-technologies offered the possibility to push the particle associated wave frequencies even higher. The radiation shielding is efficient when it denies the radiation access in the protected volume without the shield being damaged in time by radiation-combined effects of power deposition. The shield 400 is made by a few hundreds nm thick channeling layer 400 separated by the interface 406 from a high radiation absorption layer 402 stuck through the interface 405 to the backing layer 403. The incident radiation 407 is hitting the layer 401 in the input space of the nano-channels 404 acting as a resonator, that allows radiation to enter at various angles and drives it along the nanochannel 404 out of the structure 408 “bending” (term used in particle accelerator beam transport) it at various angles between 90-180 deg. The guiding structure 401, 404 is not interfering with the radiation coming from the opposite direction 409. The intermediary layer 402 separated by the interfaces 405 and 406 is used to apply electric current in order to switch or modify the reflection properties of the structure 401, making control a possibility. FIG. 5—Shows a main embodiment of the invention presenting the way a bent molecular wave-guide is made by using multi-layered clustered material. The high reflection material is made of several layers of molecular clusters. The input layer 500 is creating a rarefied electronic structure based on fullerenes or metallic spheres, able to create a resonance cavity to trap radiation inside, in the wave guide made by the molecular orbital 505. The incident wave 508, having the Poynting vector 509 towards the material, is reaching an input structure 505, 507 that guides it inside on a resonant path 506. The changes of the molecular distribution from the material 500 to 501 and 502 makes the molecular wave-guide turn, driving the wave back outside by slight interaction with electronic charges 514, and being resonantly trapped into the middle of the wave guide isopotential made of the electronic orbital surfaces 504. The material 501 is implanted over the interface 500-502 modifying the cluster end in order to make the wave-guide bend driving the wave or particle trapped inside 515 towards outside 511 but traveling in opposite direction 510. The material 502 is deposited on a structural resistance material 503 that may be a polymer, fabric or metal-ceramic sheet. The total depth of the channels remains small in the domain of 50 nm to several hundred nanometers making that the total impulse transfer due to radiation direction change by multiple small grazing angle interaction to be taken by several thousands atomic structures making the total energy taken from the radiation to be small. The slight interaction with the electronic structures and hard interaction with the nuclear structures keeps the wave on track changing the direction in small steps. The generation along the channel of orbital magnetic moments is welcomed for neutrons “gyration”, creating a focusing-defocusing molecular structure similar to particle accelerators. The resonance between neutron spin, turning and the magnetic orbital moment alternation is generating the steering force. For practical reasons, a 90 deg. gyration seems enough to comply with most of the shielding purposes. For nuclear reactor control purposes up to 180 deg. gyration angle seems appropriate to keep the radiation in a specific zone. FIG. 6—shows another practical approach of the radiation funneling mechanism by channeling into molecular wave-guides and turning it little by little. The higher input acceptance solid angle is used to trap particles inside the nano-channel and to release them through a termination with unidirectional escape angle from the channel, which will improve the previous structure input radiation admittance angle. The radiation 605 having a broad energy spectrum enters in the reflective solid 600 at a certain incidence angle and acceptable input distribution solid angle 607, accepted for channeling in the first layer 601. This layer turns the radiation a little bit and delivers it at an angle, matching the next layer 602 admission criteria. The incident radiation coming in the 602 layer acceptable angle is off the acceptance angle and did not interact by channeling with the upper layer 601, and instead adds in the 602-layer entry level with 601-output radiation. The second layer's 602 output is cumulating with the radiation coming directly inside the admission cone up to the angle limit 608 and is channeled to the nano-structure output. This way, the radiation incident on the material's surface between angles 606 and 608 forming the angular acceptance 609 is delivered inside the narrow exit angle 605. The material may be continued with several other layers in the direction 604 such as a Pi (180 deg.) incidence angle to be diverted in a single direction. FIG. 7 shows another embodiment of the invention that makes the X gamma and neutrons radiation switch from a direction to another by using a combination of ferro-electric and piezo-electric clustered materials. This is important because it introduces the capability of applying an electronic control over the direction the radiation is driven. The application of the voltage on the active layer modifies the structure and orientation of the molecular wave-guides due to electrical anisotropy of the piezo and ferro electric clusters. The simplest device is a lamellar bi-material with the piezo-electric material being deposited as a layer on the channeling material shrinking or expanding according to the control voltage. The device in FIG. 7 performs inner-intra-molecular changes at the cluster level driven by the external voltage. The nano-material is formed from the substrate 700 and the active layers 701 and 702. When the voltage is applied in one direction 705 in the adaptor module 703 the incident radiation 704 is deflected in the direction 706, shown in FIG. 7A. When the applied voltage is modified or reversed as 707 shows the incident radiation will be deflected in direction 708 as showed in FIG. 7B. The advantage is that radiation direction may be controllable being possible of making a electrically-controlled radiation shutter. FIG. 8—shows a main embodiment of the present invention called the “miu” radiation switch made of a radiation gyrator, based on a bended molecular wave-guides whose transmission is perturbed by an electrosensitive implant material controlled electronically. The material is in a planar micrometric construction having conductive layers 816 that can be used to apply the voltage 814 to control the inner channel gate mechanism. The input layer 800 is creating a rarefied electronic structure based on fullerenes or metallic spheres, able to create a resonance cavity to trap radiation inside, in the wave guide made by the molecular orbitals 805. The incident wave 808 having the direction (Poynting vector) 809 towards the material is reaching an input structure 805, 800 that guides it inside on a resonant path 806. The changes of the molecular distribution from the material 800 to 801 and 802 makes the molecular wave-guide turn driving the wave back outside by slight interaction with electronic charges, and being resonantly trapped into the middle of the wave guide iso-potential electronic orbital surfaces 804. The material 801 is implanted over the interface 800-802 modifying the cluster end in order to control the bending of the “waveguide” that is driving the wave (particle) 808 inside and turns it 815 as is going towards outside position 811 and finally traveling in opposite direction 810. The material 802 is deposited on a structural resistance material 803 that may be a polymer, fabric or metal-ceramic sheet. The total depth of the channels remains small in the domain of 50 nm to several hundred nanometers making that the total impulse transfer due to radiation direction change to be taken by several thousands atomic structures while the total energy taken from the radiation is small. The weak interaction with the electronic structures and strong interaction with the nuclear structures keeps the wave on track changing the direction in small steps. The forming along the guiding channel of orbital magnetic moments is welcomed for neutrons gyration creating a focusing defocusing molecular structure similar to those “FODO” structures used in particle accelerators storage rings. The resonance between neutron spin turning and the alternating magnetic orbital moment is generating a turning force. For practical reasons a 90 deg. gyration seems enough for most of the shielding purposes. For nuclear reactors control purposes a gyration angle up to 180 deg. seems appropriate to keep the radiation in a specific zone. The application of the voltage over the piezoelectric or ferroelectric structure makes the switch of the channels from that turning around 806 to that making a forward transfer 817 allowing the radiation to pass through without attenuation. The control voltage 814 may be applied in digital or analog manner. FIG. 9—shows a single layer of “miu” switch schematics in order to clarify its operation mode. The material is at the minimal approximation tri-layered. The intermediary layer 902 facilitation the switch 905 to 915 operation and it is separated from the input layer 901 by a conductive interface 906 that is applying the voltage between 901 and 902 layers relative to the backing layer 903 grounded by the plot 914. The voltage applied on 902 layer by the plot 912 induces a displacement in the switching structure 905, 915 such that all the parallel channels initially opened to “gyration” on the path 909 are moving to narrow the “gyration channel” while opening the direct pass channel 910. The radiation, such as gamma, neutron, X, that is coming from the 907 direction is entering the admittance resonant chamber 907 and is going to have a complete gyration along 908 direction to 904 exit chamber. When the voltage between 906, 912 plots is modified with reference to 914 the particle goes into the transmission channel 910 exiting in forward direction through 911. This represents the development of a voltage controlled variable “albedo”, shielding reflector. FIG. 10—shows the voltage-controlled reflectors, an embodiment of the present invention, application in the nuclear reactor 1008 control and shielding 1002 using the “miu” switch variable transmission 1000 and, respectively, a passive “gyrator” structure 1002. This is an important stage of the invention as allows for the drastic reduction of the shielding dimensions and mass. The nuclear reactor structure 1008 is operating using a direct nuclear energy conversion into electricity module 1008 delivering directly the electric power at the plots 1009, 1010 and using a thermal conductivity heat extraction for which the plots 1009 and 1010 represents the cooling agent exit and input in the reactor critical volume. The criticality is controlled by adjusting the neutrons transmission through the nano-shielding 1000 such as to maintain the required power level. When the nuclear reactor 1000 power has to be decreased the transmission through the shield 1000 is increased. The released neutrons are used for the fuel breeding in the blanket 1001 or fission products burning in the blanket 1006. The control shielding may have many plots to control the shielding 1011 at the local or sector level, adjusting the number of neutrons that are escaping in the outer layer blanket 1001, and separately other control plots 1012 to adjust the neutron escape flux in the top blanket 1006. The escape neutrons from the reactor's core are hitting the active shielding 1000 in the point 1003 and depending on the applied voltage applied on the plots 1011 the neutrons can reflect back in the reactor or may pass through and interact with the breeding fuel 1001 being absorbed in the point 1005 or being scattered, hitting further the nuclear reactor external shield in the position 1004 where they are reflecting back. The outer shield 1002 may be funneled by placing them tangent to breeding structure or being allowed to go along a radial path back inside the nuclear reactor 1008. In the upper side the escaped neutrons may reach the shielding 1000 in the point 1014 and, depending on the voltage applied on plots 1012, they may pass through and enter in the fission product transmutation blanket 1006 interacting in the point 1013, or may be reflected back towards the nuclear reactor core 1008. The entire structure is introduced in an external case 1007 with multiple functional roles. These new materials and procedure represent an important advancement because it makes the nuclear power accessible on vehicles, and few MWDay structures may be produced in volumes as small as cubes of 2 feet lateral size, powering a house or residence for several years, or a car that may be driven several years between refueling. FIG. 11—shows another application that uses a radiation funneling shielding device to shield a plasma focused fusion device being an embodiment of the present invention. The figure shows an application related to collider fusion device based on two beams coming from opposite direction making the center of mass of the colliding particles be in repose relative to the center of the energy harvesting geometry. The collider may be also made using a fixed perpendicular thin target and an energy harvesting structure placed after the target along the beam direction. In the figure, the role of the nano-shield 1102 is to funnel the fusion reaction secondary radiation (alpha, proton, neutron, electron) 1104 coming from the harvesting element 1103 such that to maximize its path in the absorbing element 1105 until it hits the outer nano-shielding 1101, which reflects it back tangentially 1106. The fusion assembly 1100 may have the fusion reactions of Boron 1108 with proton 1107, giving a Helium particle and a 8Be that instantly decays 1109 in two Helium particles, or Lithium 1111 with deuteron 1112 or proton, giving two Helium 1110 particles. The Helium particles carrying the fusion energy as kinetic energy interact with the direct conversion structure 1103 that takes their energy and transforms it into electricity. FIG. 12—shows another application as embodiment of the present invention; the application of the nano-layered funneling shield as a spacecraft radiation-shielding device for charged particles and X, gamma rays. In this case, the outer frame structure 1200 that protects the payload inside is shielded laterally by nano-foils 1201. This foil is funneling the radiation 1204 by controlled tunneling reflection 1203 to the energy harvesting elements 1202. For charged particles 1205, specialized inertial spinning magnetic coils 1206, are driving the charged particles into specialized or universal energy harvesting elements 1202. In this way a cosmic ray protection similar to the earth combined effect of magnetosphere and atmosphere may be achieved. FIG. 13—shows another embodiment of the present invention that is a multi-focusing multi-layer device operating as gamma, X, n, charged particle imager. Various layers 1301 having narrow admittance angle and narrow directive output create the material 1300. A radiation wave coming from the direction 1302 is transmitted on the direction 1303 towards a point in space 1304 where a detector is placed. The same happens to radiation 1305, transmitted on the direction set 1306 towards the point 1307 with the appropriate detector is placed. The radiation 1308 is also transmitted on the directions set 1309 towards the point 1310 into the appropriate detector. The radiations 1302, 1305 and 1308 may be different and come under the same incidence angle, or may be the same, and the detectors may be different. FIG. 14 shows a space-imaging device for terrestrial and outer space applications. It is mainly a 2π-3D solid angle imaging device, with angular selectivity. A plurality of such devices may generate a 3D image in spherical coordinates. The sensitive cylinder or prism 1400 contains a multitude of planar or bent layers as shown in FIG. 13, concentrating the radiation to imaging points set on the hyper-surface 1401, equipped with the appropriate transducers, with some symmetry and access shape 1402. A ray 1404 is directed inside the set angles 1405 to a specific concentrator/imaging point. The same happens with the ray 1406 poised on the direction set 1407 towards the detector 1403. The shape, detector types, radiation selectivity criteria are construction details and are subject to customization. FIG. 15—shows how an adapted nano-tube wave-guide works as an example for the buildup of the molecular wave-guides. The material 1500 is made of a plurality of layers from which the figure shows only three of them. The admittance-exit layer 1501 is followed by the channeling-in nano-structure layer 1502, placed on a substrate layer 1503. A control voltage may be applied between the extreme layers or control layers 1501 and 1503, having as effect the displacement of the atom 1507 that interferes with the admittance path, and resonator input wave trapping device 1506 made from a nano-cluster of various materials and various geometries. The incident radiation coming from the direction 1511 or 1512 is trapped in the structure 1506 and injected in the nano-wave-guide 1508. The radiation interacts slightly with the nuclei in the nano-tube that are seen at the grazing angle, being driven with almost no energy exchange towards the exit device 1504. This device matches the radiation determining its direction 1517 and the “cardioid's” 1518 shape (curve amplitude or probability of occurrence as function of exit angle) or exit angular distribution. The admittance “cardioids” 1510 are determined by the input adapting structure 1506 that makes the oscillation inside the nano-cluster 1514 adapted to be injected 1515 in the molecular guide. After channeling inside the structure 1505 the wave gets into the exit adapter 1504 having a matching 1516 oscillation before departure. It is possible that, passing through this structure, the shape and energy of the photons (it has been experimentally proven that photons have a finite number of oscillations, in the range 103 to 107, and the amplitude distribution along the string (shape) varies as function of their creation, interaction and propagation) to be modified. The selectivity between the rays X 1511,X′ 1512 and X″ 1517 is a constructive detail. The structure is reversible if the input and output matching structures are properly arranged. FIG. 16 shows the schematic view of a complete shielding structure section 1600. This is made of a plurality of layers 1601-1606 with various functions as denial or control of the radiation 1607 propagating towards 1608 the shield. The first layer 1601 may be used to harvest the energy from low penetrating radiation like charged particles up to MeV domain and electromagnetic field with energy less than few eV, delivering the energy as electricity at the plots 1611. It also acts as a protection for the next layer, depleting the radiation 1610 of its low penetration components, and may work as anti-chemical protective layer, too. This layer is not the subject of the present invention. The next layer, an embodiment of the present invention, 1602 is made of a plurality of layers containing nano-tubes or organized nano-clusters 1621, 1622 adjusted for various particles and various angles and may be controlled by the voltage applied to the plots 1623. In this layer, the radiation 1610 is “back-reflected” 1624 by “gyrating” (turning) it inside the molecular wave-guides leaving a small amount of it to be transmitted 1625 through the layers. The layer 1603 has mainly separation and mechanical resistance functions. The layer 1604 is redundant, making a supplementary gamma and n radiation absorption based on atomic absorption enhanced by nuclear resonance cascade as described in FIG. 3. The absorptive layers 1641, 1642 have various nano-micro layers of various isotopic enriched materials, which allows them to resonantly absorb the passing-by unidirectional radiation and reemit it in a broader solid angle, and by this mechanism to eliminate or dim the resonant bands from the incident radiation (gamma, n) 1643 spectrum. The layer 1606 may additionally be used redundantly as the last resort of protection, being based on mass absorption in degradation lattice as presented in FIG. 2, being mainly a usual shielding and not the object of this invention. The remnant radiation 1660 is supposed to be very low, with orders of magnitude. The dashed line 1605 is a symmetry line for the case when the shielding arrangement is bi-directional. The symmetry line may be also build on the mechanical support layer 1603 for nuclear radiation control applications. FIG. 17 shows another embodiment of the present invention related to the mechanism of atomic level radiation channeling through an atomic structure presented in a window 1700. The picture shows several atoms 1701 connected through chemical bounds at the 2p 1714 orbital level, like Carbon, which have been chosen in order to provide a simple example. The lattice might be a cluster, a nano-tube that has a specific cell dimension. The atomic electric potential curve 1702 starts at the nucleus 1701 where it has a high value decreasing fast with the distance, and being partially shielded by the electrons placed on atomic orbital 1s, and 2s 1721, 2p 1711 and 2p-bound 1714. The atomic “channel” given by the atoms alignment 1704, is bent left by a gap 1705 determining the radius of curvature of the structure by using the equation: α = Δ ⁢ ⁢ y A = A R ⇒ R = A 2 Δ ⁢ ⁢ y If keeping a smaller than 2 degrees for an interatomic distance of 3 Angstroms=A, (sp2 bound in CNT is 1.41 A) we get a radius R=20 nm. Of course this looks very small but is the lower limit a molecular wave-guide effect may occur. In reality the radiation wave 1703 has a finite length of several [nm] up to hundreds of [nm], depending on the production source, with a E; B profile wearing the signature of the primary source and the environment it passed through. In our example it has also a width and an envelope 1732 with the Poynting vector Y 1731 centered in the channel. The image resembles a ship in a strait. In normal environments between 250 and 400 Kelvin degrees the atoms have molecular vibrations at THz frequencies. Though the atoms have not fixed locations as figured by the alignment axes 1704, but likelihood places figured by the rectangles 1706 where in plain they describe a combined oscillatory movement similar to Lissajous trajectories, under the action of the figured in plane oscillations 1717 on z axis, 1718 on y axes and 1712 on XZ. In reality these movements have to be treated in volume and a plurality of specific eigen-frequency in THz domain, specific to all molecular vibrations. These movements may make the wave-guide impractical above a certain temperature, because the atoms may interpose with the wave driving to a nuclear collision effect known under the name of Doppler broadening. This effect generates Compton recoil electrons 1708 that stops far in the lattice by generating a cascade 1709 accompanied by X rays, and energy and direction modification becomes uncontrollable. This imposes the following requirements for the guiding structure: The nanowire to be straight and long, free of sudden curls, while the gap on the rotational axis to be a rational number so during a twist around its symmetry axis the molecules to cover all the space. The nano-structure has to be as compact as possible and with high electron density, such as the fields constrain to be big enough “to bend” or stir the radiation wavelet and keep it inside the channel. There is possible to vary the isotope in such a manner to create a funnel and control the exit of the particle with its associated wave from the channel. The chemical stability and the molecular strength have to be high so that the amplitude of the molecular vibration to be small enough to require no cryogenics. The developed nano structures have to be made as compact as possible to stir radiation with wavelength shorter than their interatomic distance. FIG. 18—shows another embodiment according to the invention application to build a radiation selective extractor/gyrator 1800. A radiation beam 1802 of composed radiation reaching the target 1801 may be separated on types of incident radiation that may be further extracted from the “hot area” by tuned guiding tubes 1804, 1808, driving it to different receivers. The radiation may be a mixture of n, gamma (neutrons decay becoming p, e, for travel times greater than ½ hour due to n disintegration). It is possible to develop a communication system that relies on the n emitter modulation, ovelaped with gamma that travels with different speeds being necessary at receiver to separate the gamma from the rest of the radiation in order to get the desired signal. Their overlap on target receiver makes the decoding hard due to physical properties of the signal that have to be extracted from the high-energy radiation background. This kind of communicator is also usable in high radiation environment where the noise 1805 may be distinctly extracted and separated from the real information-carrying signal 1807. The system is transparent to the radiation that is not matching the extraction conditions for the path 1806. The nano-structured entry interface 1801 takes all the radiation that is falling on receiver 1803 to the input of the specialized extraction guides acting like filters. This device may use the signal decoders for imaging and communication purposes. The radiation modulation might be done with the electro-sensitive radiation transport device shown in FIG. 9. In FIG. 19 is shown another possible application of the thin radiation-guiding based shield for a whole body protective coat 1900, formed by the upper body coat 1907 and pans 1908 with protective shoes or boots 1909. The coat 1901, 1907 may include helmet or hood 1903, and a backpack 1902 for survival and instrumentation. The face protection may have a face protection shield 1904 that may be transparent for eyes or completely opaque equipped with complex orientation system 1905, giving the images of the terrain in various bands and radiation. The gloves 1906 may have various degrees of flexibility and protection. The advantage of this suit is that it may exhibit attenuation coefficients up to ppm level and weight by 100 times less than it would if being fabricated by the current technology using mass-attenuation based materials. Some flexibility will be possible in the suit. The suit may be used in various configurations and circumstances for individual protection such as military suit, security first intervention, hazmat environments, outer space for astronaut suit or for outposts shielding, for shelter in place, portable emergency vehicles, etc. FIG. 20 shows a complex application of these active shields 2000 in making lighter and smaller the nuclear power source, like fission, fusion and hybrids nuclear reactors. The entire assembly is contained into a technologic case 2001, built for mechanical resistance and shielding purposes, able to contain pressure and heat. The total reflection external shield 2002 is reflecting all the particles coming from the central hot area back towards inside, keeping the radiation together in a small confinement zone. Immediately near the absorber is placed a multipurpose cooled absorber material 2003 making a sealed structure able to confine pressure and heat, up to a limit where it has a controlled release. A direct conversion layer and a gamma absorber material form the absorber 2003. It also contains fission products chemical stabilization materials. The next containment layer is used for breeding control being an adjustable reflection shield 2004 that is driving the neutrons to the absorption layer 2003 or towards inside in the fertile blanket having the function of controlling breeding and transmutation rate. When the structure is electrically polarized, is acting as a reflector driving the escaped particles back inside to be used for breeding or power production, while is transparent and leaves particles escape in the absorption blanket 2003. The breeding nano-structured-material 2005 contains 232Th, 238U, or may contain other materials for radioisotopes production by transmutation. In order to maintain a constant reactivity in this area, the transmutation products will be removed from the production area in a storage/cooling compartment outside the active zone. The reactor's power control adjustable-reflection shield 2006 is made form the active “miu-switch” material, whose transmittance and reflection (also called albedo) is adjusted using an electronic control system for criticality 2007, that together with the breeding control system 2008 assures the neutron 2009 flux management, establishing their trajectories 2010 and their range. All the functions are automatically controlled in order to balance the power output at the plots 2011 to meet the power demand. FIG. 21 represents an example of a mobile, portable nuclear power source 2101, based on reflective, active nano-shielding 2102 that may lead to advanced nuclear sources, like for example a 239 Pu fueled nuclear reactor reaching the criticality with less than 50 g of fissile material. The produced power is transmitted to electric motors 2103 in the wheels that makes part of the integrated vehicle power system 2100. Other examples of mobile applications are trains, ships, planes, super-planes space shuttles and underwater devices. The present invention refers to a new type of active nano-structured material to be used for X, gamma and neutrons shielding that may have the capability to electronically control the flux reflected or transmitted through it. The patent is a development from the current super-mirror used in synchrotrons X ray focusing and cold neutron transport at spallation sources using active nano-structures. Another idea used in the patent approach relies on the fact that the interaction between high-energy radiation and materials is very weak except for nuclear resonances. Such resonant materials may have small thickness but may generate high absorption rates. The equation 1 is characterizing the classical mass absorption process, including resonant materials:I(Ej)=I0(Ej)exp(μixi+δjiμjxj) Eq. 1where I is the intensity in a point x on the axis for an energy belonging to the energy group j, and i is an index counting the attenuator materials in the shield. The “Kronecker delta” is suggesting that only for the resonant energies an extra term may be added to count for resonant absorption and reemission of radiation similar to scattering. The energy group “j” width is set to be equal with the resonance's effective width (something like nσ where n is a reasonable value usually smaller than 3 and sigma is the distribution parameter). μ is the linear absorption coefficient absorbing the value μ ρ ⁢ ρ i where μ is the material specific absorption density while ρ is the material mass density and ρi is the specific material density spanning the length xi. Using this concept there is possible to make arrangements of various materials resonantly absorbing the incident radiation, activating the internal nuclear channel and de-exciting by following the nuclear branching paths. There are very few cases when the excited nucleus is emitting a higher energy than it absorbed, therefore the new material becomes a new source of radiation in that bandwidth backscattering theoretically 50% of the primary radiation. If consider two repetitive layers separated by a distance they theoretically cut down 75% of the radiation by backscattering. The disadvantage of these materials is that the resonance band is very narrow, therefore a sandwich is required to cut down most of the energetic groups. The nature did not provide so many stable isotopes as we may need to make an efficient resonant shielding for high energies. Radiation buildup is also important but is considered a secondary effect for this approach. This is what is already known and serves as background for this invention. The usage of the concept of radiation reflection at grazing angles together with the fact that the radiation interaction with the surface is local, involving few tens of atoms, led to the conclusion that a nano-tube slightly bending, see FIG. 17 may offer the same conditions, and more, it may contain the radiation inside and slightly centered in the tube. The nano wires are naturally bending and so does the radiation trapped inside. One problem to be solved is the small admittance for trapping radiation inside the molecular wave-guides, which is solved by growing structures like fullerene that will divert the radiation inside, as described in FIG. 15. FIG. 1 describes 3 types of shielding. The novelties brought by this patent are the nuclear resonance enhanced absorption and the molecular wave-guides radiation “gyration”. If the resonance enhanced radiation absorption and reemission described in FIG. 3 brings a passive shield with maximum 1-2 orders of magnitude thinner than the classical mass absorption current shielding described generically in FIG. 2, the new radiation gyration by molecular wave-guides described in FIG. 4 opens new perspectives. As FIG. 5 shows, there is necessary to build organized nano-structures, in order to create the so-called molecular-guide structures. These structures may be build in many ways, but for simplicity, one way to build is by starting from a Si or Diamond substrate, building by beam-annealed Au self organized nano-clusters, and building a layer of carbon nano-tubes, slowly bent in about 500 nm to 1 micron. Over this layer a new conductive micro layer is deposited as TiO and W, or WC follows by pulsed laser deposition than by Au, Ag deposition. This substrate will create the germination for the new set of C nano-tubes deposited by CVD, slightly tilted than the first. In this way a plurality of substrates may be build. Another modality of building the organized structure is to perform a combined CVD and Laser Pulsed Deposition, assisted by an interfered ion bean on a 10 nm pattern to create the thermal spikes to induce the nucleation of the nano-clusters and separation of the depositions. In this structure the organized layers of nano-clusters will float in an insertion material, also partially crystallized. The insertion of a piezo material as BaTiO4 by LPD or a ferro-electric material as TGS brings the possibility of the electric control of the radiation direction by obtaining the molecular wave-guide switches. As already resulted from FIG. 5 and in the molecular switch version presented in FIG. 8 there is difficult to achieve this effect by using few atoms. A collective action with the participation of few thousands atoms is needed to completely gyrate neutrons with energy up to few MeV or gamma rays as detailed in FIG. 17. In the case of the gyration by 180° of the radiation of few MeV on 1000 atoms, an energy exchange of several tens of eV will be transferred to lattice due to momentum transfer. This is enough to warm-up that channel and the structure to require cooling. The formula is: p = 2 ⁢ E c = nMv → v = 2 ⁢ E cnM ⇒ Δ ⁢ ⁢ E = 2 ⁢ E 2 c 2 ⁢ nM = nk B ⁢ T ( 3 ) that in the case of 1 MeV radiation turned by 1000 atoms gives about 2.5 eV, leading to a 30K equivalent temperature increase per particle. This is not so bad showing that high doses may be handled by this mechanism without significant radiation damage effects. To calculate the radiation damage the isotopic specific interaction cross-sections have to be considered. Without doing this, we observe that in the radiation admission interface small cross-section materials have to be used to channel the radiation inside the nanostructure. The particularity of the channeling process exploited in the present invention consists in the fact that the radiation quanta interacts mainly with the collective atomic electric field and not directly with the nuclei, making the interactions smaller than in the case of nuclear interaction The application of this material in communications applications, as shown in FIG. 18, uses the selectivity and electric control that makes possible the modulation of the emitter and the use of direct radiation energy conversion in electricity systems, having a fast response necessary for detection and demodulation. The high sensitivity of detectors and low absorption of the neutrons and gamma rays in materials makes possible the signal transmission through shielding materials and high radiation absorbers. The usage of these active radiation guiding materials inside a nuclear reactor is making possible the replacement of the mechanical control rods by electrically controlled “albedo” materials, increasing the neutrons usage and making an optimal management of breeding, transmutation and partitioning processes. The nuclear waste and structural materials contamination will be drastically reduced. This will lead to major changes in the nuclear reactor structure and design. Same active radiation nano-guide structure might be used to enhance fusion structures, accelerator driven nuclear reactor structures and hybrid nuclear reactor structures.
	abstract	A reflection surface 12 constituted by a transition metal having a core level absorption edge in the vicinity of a wavelength of a soft X-ray is formed on an inside of a vacuum vessel 14, and furthermore, there is provided a permanent magnet 13 for generating a magnetic field in a perpendicular direction to a longitudinal direction of the vacuum vessel 14 in a position of the reflection surface 12 by which the soft X-ray is to be reflected, and the soft X-ray to be linearly polarized light incident on the vacuum vessel 14 is reflected at plural times over the reflection surface 12 in a position where the magnetic field is applied in such a manner that magnetic scattering is increased by a resonant effect of a magnetic circular dichroism when the soft X-ray is reflected by the reflection surface 12. Thus, a great difference in a refractive index is made between circularly polarized counterclockwise light and circularly polarized clockwise light which constitute the linearly polarized light, and a phase difference between the circularly polarized counterclockwise light and the circularly polarized clockwise light is obtained at a time. Consequently, it is possible to reversibly convert the soft X-ray from the linearly polarized light into the circularly polarized light or from the circularly polarized light into the linearly polarized light by a reflection to be carried out at only several times.
051401657	summary	BACKGROUND OF THE INVENTION This invention relates to a solidification treatment vessel into which radioactive waste pellets are charged, and then a solidifying material in the state of a liquid or a slurry is poured into the vessel to solidify the radioactive waste pellets. The invention also relates to a method of solidification treatment of such pellets, using the above vessel. Recently, there has been used a method of treating a radioactive waste in which a concentrated waste liquid, a waste sludge, a used ion exchanger resin or the like is formed into powder by a vertical thin film dryer, and then the powder is formed into pellets of the tablet type or the briquette type by the addition of an appropriate amount of a suitable binder. There has also been used a method in which ash resulting from combustible wastes burnt in an incinerator is suitably sieved, and then is formed into pellets in the same manner as described above. Such pellets are charged into a suitable vessel or container, and then a liquid-state or slurry-state solidifying material of an organic or an inorganic type (e.g. a cement type or a plastic type) is poured into the vessel to uniformly fill the spaces between the pellets, and is solidified in the vessel, thereby forming a stable solid mass. In the above pouring of the solidifying material, if the specific gravity of the pellets is lower than that of the solidifying material slurry, the pellets tend to float on the surface of the solidifying material slurry at a final stage of the pouring operation. Therefore, without a suitable float prevention means, many pellets float and appear on the surface of the solidifying material. In one conventional method of preventing such a float phenomenon, after a predetermined amount of pellets are charged into the vessel, an iron frame or the like having a metal net whose mesh is smaller than the size of the pellet is fixed to the upper portion of the vessel, and then the solidifying material slurry is poured into the vessel through the metal net from above this metal net, thereby preventing the floating of the pellets. Then, the vessel is left to stand until the viscosity of the solidifying material increases to such a degree as to restrain the floating of the pellets, and then the iron frame with the metal net is removed from the vessel, and then a required amount of solidifying material is post-filled in the upper space of the vessel. This conventional method has the following disadvantages: If the iron frame with the metal net is fixed to the upper portion of the vessel before the pellets are charged into the vessel, a subsequent charging of pellets into the vessel is prevented by the metal net. Therefore, in the above conventional method, the iron frame with the metal net is fixed to the upper portion of the vessel after the pellets are charged into the vessel, and then the solidifying material is poured into the vessel. Thereafter, the iron frame with the metal net is removed for re-use. However, after the charging of the pellets, the wall of the vessel is stained with dust of the pellets, and the iron frame with the metal net can not be easily attached to the vessel. And besides, a mechanism for the attachment and detachment of the iron frame relative to the vessel is complicated, and an automatic remote control which is desired for preventing the radiation exposure can not be easily achieved. In addition, when the iron frame with the metal net is to be removed from the vessel after the pouring of the solidifying material, the solidifying material is liable to be scattered over the surroundings, and it is troublesome to handle the thus removed iron frame. To avoid these problems, it can be considered that the iron frame with the metal net is not removed from the vessel even after the pouring of the solidifying material, and is embedded in the solidifying material in the vessel. In this case, also, the iron frame with the metal net must be attached to the upper portion of the vessel after the pellets are charged into the vessel. Therefore, the same problems as described above are also encountered. In order to charge a Predetermined amount of the pellets into the vessel with preventing the jumping of the pellets out of the vessel and the scattering of the dust, it is preferred that a kind of lid be mounted on the upper portion of the vessel and that the pellets are charged into the vessel via a charge pipe extending through this lid. However, even with this pellet-charging method, it is impossible to charge the pellets, with the iron frame with the metal net (which prevents the floating of the pellets when the solidifying material is poured at a later stage) being beforehand attached to the upper portion of the vessel. Therefore, after the charging of the pellets, the lid is removed from the vessel, and thereafter the iron frame with the metal net must be attached to the upper portion of the vessel. Thus, the same problems as described above are also encountered. SUMMARY OF THE INVENTION With the above deficiencies of the prior art in view, it is an object of the present invention to provide a vessel for solidifying radioactive waste pellets which vessel enables the charging of the pellets and the pouring of a solidifying material without the need for attaching and detaching a pellet float-prevention member, and also enables an easy automatic remote control of such charging and pouring. Another object of the invention is to provide a method of solidifying radioactive waste pellets, using the above vessel. According to one aspect of the present invention, there is provided a vessel for the solidifying radioactive waste pellets comprising: a vessel body; an inner lid mounted within the vessel body and fixedly secured to an upper portion of the vessel body, the inner lid having an opening formed at a generally central portion thereof; and means for preventing the pellets from floating fixedly secured at one end thereof to the inner lid and extending into the opening to define therebetween gaps allowing the passage of a solidifying material in the state of a liquid or a slurry therethrough but preventing the passage of the radioactive waste pellets therethrough, the pellet float prevention means, when receiving a downward urging force, being bent downward to enlarge the gaps for allowing the radioactive waste pellets to pass therethrough, and the pellet float prevention means being returned by a resilient restoring force to its initial positions when the downward urging force is released. According to another aspect of the invention, there is provided a method of solidifying radioactive waste pellets comprising the steps of: (a) providing a vessel for solidifying the radioactive waste pellets of the above-explained; (b) lifting the vessel so as to insert a charge pipe through the opening of the inner lid, so that the pellet float prevention means is bent by the charge pipe; (c) subsequently charging the radioactive waste pellets into the vessel through the charge pipe; (d) subsequently descending the vessel so as to remove the charge pipe from the opening, so that the pellet float prevention means is automatically returned to their respective initial positions; and (e) subsequently charging a solidifying material into the vessel while preventing the floating of the charged pellets by the pellet float prevention means, and allowing the solidifying material to solidify, the solidifying material being in the state of a liquid or a slurry and being higher in specific gravity than the pellets. In the present invention, the pellet float prevention means is bent to enlarge the gaps therebetween so as to charge the pellets into the vessel body through these gaps. After the charging of the pellets, the pellet float prevention means is returned by the resilient restoring force to its initial position, so that the gaps between the pellet float prevention means do not allow the passage of the pellets therethrough. After the pellet float prevention means is returned to its initial position, the solidifying material in the state of a liquid or a slurry can be poured through the above gaps, while preventing the floating of the pellets by the pellet float prevention means. Therefore, the charging of the pellets and the pouring of the solidifying material can be carried out quite easily, as compared with the prior art in which the iron frame with the metal net must be detached and attached before and after the charging of the pellets. The pellet float prevention means of the bendable, restoring type are provided at the opening of the inner lid, and the pellets are charged by inserting the charge pipe through the opening of the inner lid. Therefore, the scattering of dust and the jumping of the pellets out of the vessel at the time of the charging of the pellets can be prevented. Further, the inner lid can be embedded, together with the pellet float prevention means, in the poured solidifying material, and do not need to be removed, and therefore the solidifying of the pellets can be carried out more easily.
046506322	abstract	In a magnetic plasma confinment device having an inner toroidal limiter mounted on an inner wall of a plasma containment vessel, an arrangement is provided for monitoring vertical temperature profiles of the limiter. The temperature profiles are taken at brief time intervals, in a time scan fashion. The time scans of the vertical temperature profile are continuously monitored to detect the presence of a peaked temperature excursion, which, according to the present invention, is a precursor of a subsequent major plasma disruption. A fast scan of the temperature profile is made so as to provide a time interval in real time prior to the major plasma disruption, such that corrective action can be taken to reduce the harmful effects of the plasma disruption.
	abstract	This leakage prevention seal is provided with: a first seal ring which surrounds a rotating shaft and which is in circumferential contact with the surface of a housing which faces the upstream side; a second seal ring, which, on the upstream side of the first seal ring, surrounds the rotating shaft and which is in circumferential contact with the first seal ring; and a heat-driven section which, when high-temperature pressurized water reaches the heat-driven section, reduces the diameter of both the first seal ring and the second seal ring and causes the inner peripheral surfaces of both the first seal ring and the second seal ring to be in contact with the rotating shaft. The circumferential positions of both a first range between the first seal ring and the rotating shaft and a second range formed by the second seal ring are different.
	description	This application claims priority under 35 U.S.C. §119(e) to U.S. Provisional Patent Application Ser. No. 61/715,367, filed Oct. 18 2012, entitled JET PUMP REPAIR FOR A NUCLEAR POWER PLANT. 1. Field The present invention relates to a method for repairing a jet pump and more particularly to a method for repairing the slip joint between an inlet mixer and a diffuser of a jet pump with particular benefit to jet pumps employed in boiling water reactors. 2. Related Art As can be appreciated from FIG. 1, in conventional boiling water reactors 10, jet pump assemblies 18 are located in the reactor vessel's annulus region 12, between the core shroud 14 and a wall of the reactor vessel 16. The primary function of the jet pump is to pump coolant below the reactor core where the coolant then flows up through the fuel assemblies to extract heat from the fuel assemblies. The jet pumps are also relied upon to maintain two-thirds of the water level height in the reactor core during postulated accident conditions. For these reasons, the jet pump assemblies 18 are considered safety related components. In a typical arrangement, twenty jet pumps are paired in ten assemblies 18, as illustrated in FIG. 2. Each jet pump assembly 18 is driven by flow from a common riser pipe 22. Jet pump flow is then directed to the lower plenum region 24 below the core supported within the shroud 14. Jet pump flow is roughly one-third driven by the reactor coolant system pump flow through the riser pipe 22, and the last two-thirds of jet pump flow is due to venturi action of the jet pump suction inlet that pulls in fluid from the annulus region 12. The jet pump assemblies 18 are circumferentially spaced around the shroud 14, supported on a shroud support ledge 20 near the bottom of the shroud. Industry operating experience has encountered numerous instances of damage or accelerated wear to critical components of the jet pump assemblies 18 which has affected a large population of boiling water reactors both in the United States and globally. The most common jet pump components to experience damage are the main wedge and rod 26, restrainer bracket pad in the restrainer bracket 28, riser pipe 22 welds and riser brace 30, all of which components can be observed in FIG. 3, which shows an enlarged perspective view of a jet pump assembly 18. These common types of damage are all in critical supporting structures or features of the jet pump assemblies. Damage and wear are largely attributable to flow induced vibration due to normal operation. In addition to flow induced vibrations, many plants experience excessive leakage in the slip joint region 32, which can exacerbate the vibration experienced in the jet pump assemblies 18. This leakage can greatly accelerate the degradation and damage done to the jet pumps. Plants that have shown signs of accelerated wear and component movement, both indicators of flow induced vibration issues, have historically attempted to address the problem by adding additional hardware to help support the jet pump components and reinforce the components against flow induced vibration. These solutions have generally been in the area of the restrainer bracket 28 and include larger main wedges 26, supplemental auxiliary wedges, and slip joint clamps. These solutions do not address the root cause of flow induced vibrations, have been ineffective for some plants, and their effectiveness overall is questionable. In addition to these typical solutions, a few boiling water reactor plants have added labyrinth seals to the inlet mixer 34 original equipment manufacturer design. These labyrinth seals are intended to reduce bypass flow in the slip joint region 32 of the jet pumps; however, it appears that under certain operating conditions in some plants these seals have been ineffective and the seal geometry has been damaged on the inlet mixer's outer diameter and has caused damage to the inner diameter of the collar 38 on the diffuser 36. In the slip joint region 32, the inlet mixer 34 is unsupported or floats, allowing the mixer to thermally expand along its length during plant startup and shutdown. The inlet mixer's bottom section fits into the collar 38 of the diffuser assembly 36, forming a slip joint. The inlet mixer 34 is laterally supported by a three-point contact at the restrainer bracket 28. This three-point contact is maintained with a sliding (main) wedge 26 and two set screws that are tack welded in place. The main wedge 26 is held in place by gravity, theoretically resulting in three-point contact. The very upper portion of the inlet mixers are supported by a pre-tensioned beam bolt assembly 40 that presses down on the inlet mixer 34 where it is seated in the transition seat 42. Because of how the inlet mixers 34 are supported, small lateral loads on the bottom of the inlet mixer (within the slip joint 32) can create large reaction moments at the restrainer bracket 28. As previously mentioned, the main wedge 26 is held in place by its weight, typically about eight pounds, which can be overcome and lifted by small lateral forces. Since the inlet mixer 34 weighs significantly more than the wedges 26, its mass can easily overcome the holddown force of the main wedge 26 with small lateral displacements at the outlet of the inlet mixer 34 within the diffuser collar 38. Once the wedge is temporarily displaced, three-point contact is lost, and, in severe cases, the bottom of the inlet mixer may hammer against the inside of the diffuser collar 38. This hammering of the inlet mixer 34 and diffuser 36 can also be excited at particular frequencies of vibrations, potentially caused by drive flow or bypass flow in the slip joint 32. Thus, a new solution to flow induced vibrations is desired that will address the root cause of the vibrations. Furthermore, a solution to the flow induced vibration wear is desired that will minimize such wear and require little or no disassembly of the jet pump assembly 18. Further, such a repair is desired that can be performed remotely, under water. These and other objects are achieved by employing a new method of repairing a slip joint on a jet pump assembly between an inlet mixer and a diffuser that has an opening that receives the inlet mixer with a given spacing between an outside diameter of the inlet mixer and an inside diameter of the opening in the diffuser forming an annulus; with the given spacing a product of manufacture and vibration wear. The method comprises the steps of remotely accessing the annulus and narrowing a radial dimension of the annulus. In one embodiment, the method includes the step of measuring a dimension of the outside diameter of the inlet mixer that fits within the slip joint. A clamp is then fabricated having a generally circular collar clamp opening with a design diameter that is larger than the outside diameter of the inlet mixer and smaller than a maximum extent of the inside diameter of the diffuser opening. A collar clamp is then fitted around the inlet mixer and at least partially over and above the diffuser opening with the collar clamp supported by the diffuser. The collar clamp is then attached to a portion of the diffuser housing below the diffuser opening. Preferably, the measuring step measures dimensions around the diffuser opening in addition to the outsider diameter of the inlet mixer. In one embodiment, the diffuser has guides spaced circumferentially around a housing of the diffuser, with the guides extending above the opening in the diffuser that receives the inlet mixer. In the latter embodiment, the method includes the steps of forming notches in an underside of the collar clamp, in line with the guides; and fitting the notches over the guides wherein the guides restrain rotation of the collar clamp. In these embodiments, the collar clamp effectively optimizes the insertion depth of the inlet mixer within the diffuser opening. Desirably, the collar clamp is fabricated in at least two circumferential sections with each of the sections fastened together to form the generally circular opening. In this embodiment, the attaching step clamps the collar clamp to the portion of the diffuser housing, which is preferably a radially outwardly extending collar on the diffuser housing. In this latter arrangement, the collar clamp has at least two radially, outwardly extending segments that extend out radially further than the diffuser collar and the outwardly extending segments have a vertical opening therethrough. A tie bar having a radially, inwardly extending lip at a lower end positioned under the diffuser collar and a second end of the tie bar extending through one of the openings in the segments is captured on another side of the opening in the segments to tighten the collar clamp down against the diffuser collar. Preferably, the attaching step clamps the collar clamp to the portion of the diffuser housing at a plurality of discrete circumferential locations around the housing. In this arrangement, the method does not require the step of removing the inlet mixer from the diffuser. In still another embodiment, the collar clamp has an axially extending convergent surface that faces an outer surface of the inlet mixer when the collar clamp is fitted around the inlet mixer and the collar clamp rests on a lip of the diffuser opening. In one arrangement, the collar clamp has an annular circumferential groove adjacent the generally circular clamp opening, the groove having a generally “L” shape in the radial direction with one leg of the “L” extending in a horizontal direction and resting on a lip of the diffuser opening. Preferably, the second leg of the “L” contacts an outer wall of the diffuser. The method may also insert a gasket between the collar clamp and a lip of the diffuser opening to minimize leakage. In still another embodiment, the step of narrowing the radial dimension of the annulus comprises the step of removing the inlet mixer from the diffuser. Then, the inside surface of the diffuser opening is machined and material damage on the inlet mixer outer surface that is to be inserted into the diffuser opening is resurfaced. The method then inserts an internal collar having an outside diameter substantially equal to an inside diameter of the machined inside surface of the diffuser opening and has an inside diameter that narrows the annulus gap when the inlet mixer is inserted into the diffuser opening so that the annulus has a radial dimension that is less than the given spacing. Preferably, the internal collar is fabricated to have an axially convergent contour on a surface that opposes the outer surface of the inlet mixer. In each of the foregoing embodiments, the radial dimension of the annulus is narrowed to be equal to or smaller than a corresponding original equipment manufacturer specification. Alternately, in a separate embodiment, the step of narrowing the radial dimension of the annulus includes the step of cutting a collar portion of the diffuser that surrounds the inlet mixer from the remainder of the diffuser. The collar portion of the diffuser is then removed from the rest of the diffuser and the inlet mixer. A spool piece is then fabricated having a replacement opening with a desired inside diameter to replace the collar portion of the diffuser; and the spool piece is secured to the rest of the diffuser with an end of the inlet mixer within the replacement opening. FIG. 4 shows a close-up perspective view of the slip joint region 32 with the diffuser 36 having a radially outward projecting shoulder 44 just below the diffuser collar 38 that defines the opening 46 in the diffuser in which the inlet mixer 34 is inserted. The diffuser collar 38 has guides (sometimes referred to as ears) that extend radially outward and upward from the opening 46 to guide the inlet mixer 34 into the opening 46. According to one embodiment of the present invention, the current inlet mixer 34 and diffuser collar 38 are supplemented by stacking an additional collar clamp 48 on top of the diffuser, over and around the diffuser collar 38 as shown in FIG. 5. In addition to other benefits, the collar clamp 48 optimizes the overall insertion depth for the slip joint 32. The insertion depth of the inlet mixer 34 into the diffuser 36 has been recognized as one of several critical parameters that lead to the onset of inlet mixer vibration. In another embodiment, the collar clamp 48 is structured to create an axially convergent slip joint geometry relative to the diffuser and/or inlet mixer. Rather than replace the existing inlet mixer, the design of the present embodiment retains and creates a new slip joint region 32 by the addition of hardware onto the top of the diffuser 36. The present invention addresses flow induced vibration issues by either: (1) using a convergent slip joint design, (2) optimizing the effective insertion depth, or (3) both using a convergent slip joint design and optimizing the effective insertion depth. FIG. 5 shows this design concept developed according to one embodiment of the invention. The embodiment shown in FIG. 5 allows for any existing damage to both the mixer and the diffuser to be left in place. A new slip joint area is created directly above the old slip joint (see FIG. 6). This design approach has the following advantages: (i) it is able to be installed in situ with no inlet mixer removal required; (ii) does not require surface repairs; (iii) provides tight tolerance control of the slip joint gap; (iv) creates an optimal insertion depth; (v) reduces overall repair time and costs; (vi) enables a convergent slip joint configuration; and (vii) provides a flow-induced vibration solution that addresses a root cause. Generally, there are two basic options for implementing an axially convergent slip joint design onto the existing diffuser components. The first option entails modifying the exiting diffuser surface by removing or adding material on its collar 38. The second option entails adding additional hardware and creating a new slip joint area above the old slip joint area, e.g., as described with respect to FIG. 5. The present invention contemplates both options. FIG. 6 is a cross section of the embodiment shown in FIG. 5; i.e., the second option mentioned above. Bracket A on the right shows the original insertion depth prior to the collar clamp 48 being installed. Bracket B, just to the left of bracket A shows the original slip joint area. Bracket C on the left shows the new improved slip joint area achieved by adding the collar clamp 48. FIG. 7 shows one embodiment of the collar clamp 48 that is created from two semi-circular segments 52 and 54 which are joined by dovetail joints 56 and 58, though it should be appreciated that other means of joining the segments are available and the clamp 48 may be constructed out of two or more such segments. Each segment has a radially outwardly extending arm 60 and 62 through which holes 64 and 66 are formed that will be used to clamp the collar clamp 48 to the diffuser housing 36 as will be described hereafter. FIG. 8 is an enlarged partial sectional view of the inlet mixer 34, the diffuser 36 and the collar clamp 48 embodiment shown in FIG. 5, uncovering the convergent slip joint at the intersection between collar clamp segments. The design utilizes a convergence geometry 68, i.e., the inner face of the collar clamp 48 that faces the outer surface of the inlet mixer 34 converges toward the outer surface of the inlet mixer as one progresses from the upper and lower ends to the center of the inside face of the collar clamp 48. The convergence geometry works off an unmodified inlet mixer original equipment manufacturer outer surface design. The actual dimensions and angles can be fine tuned for each slip joint (since the existing slip joint geometry is left in place and new differential pressure conditions are created in the slip joint). According to one embodiment of this invention, digital measurements are taken and three D models rendered of the inlet mixer 34 and diffuser collar slip joints 32, for example, using a three-D laser scanner. These measures are taken since the as-found conditions of the diffuser and mixer may differ between jet pumps (i.e., components will vary dimensionally from one another, and actual as-built dimensions are unknown). Also, the tight tolerance for the slip joint gap requires the added hardware to have high tolerance requirement for fit-up. The three-D laser scanner technology provides very accurate measurements, approximately plus/minus 0.005 inch (0.013 cm). Also, the rendered three-D model may be saved as a compatible AutoCAD file type, which allows a machine shop to use the CAD file to automatically program CNC mills and lathes to machine from hardware blanks which meet these tight tolerances. According to the current embodiment, the collar clamp configuration uses two stack halves 52, 54 that interlock the dovetail joints 56, 58 formed at their circumferential ends; see FIG. 7. To ensure the collar clamp 48 cannot be raised up off of the diffuser upper lip, two tie bars 70, 72 clamp down on the collar clamp 48; leveraging off of the bottom edge of the diffuser shoulder 44. Each tie bar has a laterally inwardly extending projection 74 that seats under the diffuser shoulder 44 against which the tie bars 70, 72 react to maintain the collar clamp 48 pressed against the upper lip of the diffuser opening 46 (FIG. 5). The upper portion of the tie bars is threaded so that nuts can tighten down the tie bars, applying a slight preload. Preferably, these nuts are crimped in place by crushable material built in to the nut or collar clamp 48. A crushable gasket may be employed if needed between the diffuser stack 48 and the diffuser lip to ensure there is no leakage at their interface. In order to prevent the collar clamp from rotating, the diffuser guides 50 are used as support surfaces. The diffuser guides (often called ears) purpose is to help align and aid in the insertion of the inlet mixer 34 during jet pump reassembly. Notches 76 are formed in the underside of the collar clamp that allow the external collar clamp 48 to fit over the ears 50 and down onto the diffuser lip. These notches also prevent the collar from rotating. The ear recesses in the collar clamp may allow some leakage, but only small amounts of bypass flow are likely. The hardware shown in the embodiment illustrated in FIG. 5 is light enough that the two stack halves can be delivered remotely using tool poles. Much of the tooling necessary for installation exists, and minimal if any new hardware handling tooling is required. While the current embodiment illustrates one design for clamping the stack halves together, it should be appreciated by those skilled in the art that this invention is not limited to this particular embodiment. FIG. 9 illustrates another embodiment for repairing damaged jet pump surfaces and/or reducing/eliminating flow induced vibration. According to the embodiment shown in FIG. 9, damaged material on the inside diameter of the diffuser is resurfaced. FIG. 10 is a sectional view of the embodiment shown in FIG. 9. A new internal collar 78 is inserted into the diffuser collar 38, restoring it to at least its original designed inside diameter, or even narrowing the annular gap between the inlet mixer and the diffuser collar. The originally engineered manufactured tolerance gap for the slip joint between the inlet mixer outside diameter and the diffuser collar is very tight, plus/minus 0.010 inch diametrically. It should be appreciated that the new internal collar may also be structured to form a convergent geometry relative to the outside diameter of the inlet mixer. According to another embodiment of the present invention, the inlet mixer is left in place, but the diffuser collar portion 38 of the diffuser is cut and removed. A new casting or spool piece 80 is then secured to the diffuser 36 (FIGS. 11 and 12). This allows the slip joint geometry to be tightly controlled. This spool piece can be a single section (which may require the removal of the inlet mixer for installation) or multiple sections (i.e., like a clam shell) which may allow for the inlet mixer to in situ install. Again, the geometric relationship between the inlet mixer and the diffuser can be structured such that the inlet mixer outside diameter surface and the diffuser inside diameter surface converge. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
	claims	1. In a process for cleaning a radiopharmaceutical reusable pig having a lower portion, a cap, and a syringe or vial containment enclosure, comprising the steps of:scanning the pig with a radiation detection device to detect the presence of radioactivity at a radiation level above background level;upon detecting the presence of radioactivity at said radiation level, causing said pig to cease exhibiting said radioactivity; andsanitizing the pig to destroy any microorganisms and remove any blood contamination from said pig;the improvement comprising the additional steps of:transporting the ceased exhibiting radioactivity and sanitized pig to a drug preparation area suitable for dispensing a drug for human use;within the drug preparation area, inserting a syringe or vial containing a radioactive drug into the containment enclosure and assembling the cap to the lower portion of the pig;within the drug preparation area, placing the assembled pig containing the drug, in a protective outer container to protect the pig from external contamination during handling and transportation; andplacing the protective outer container containing the pig, in a transportation receptacle. 2. The improvement according to claim 1, comprising the additional step of sterilizing or sanitizing the pig after said pig is in the drug preparation area and prior to inserting a syringe or vial into the containment enclosure thereof. 3. The improvement according to claim 1, comprising the additional step of sterilizing or sanitizing the protective outer container before the pig is placed within it. 4. In a process for cleaning a radiopharmaceutical reusable pig having a lower portion, a cap, and a syringe or vial containment enclosure, comprising the steps of:scanning the pig to detect the presence of radio-activity at a radiation level above background level;upon detecting the presence of radioactivity at said radiation level, causing said pig to cease exhibiting said radioactivity; andsanitizing the pig to destroy any microorganisms and remove any blood contamination from said pig;the improvement comprising the additional steps of:transporting the ceased exhibiting radioactivity and sanitized pig to a drug preparation area suitable for dispensing a drug for human use;sterilizing or sanitizing the pig after said pig is in the drug preparation area;within the drug preparation area, inserting a syringe or vial containing a radioactive drug into the containment enclosure and assembling the cap to the lower portion of the pig;within the drug preparation area, sterilizing a protective outer container for the pig;within the drug preparation area, placing the assembled pig containing the drug, in the protective outer container to protect the pig from external contamination during handling and transportation; andplacing the protective outer container containing the pig, in a transportation receptacle. 5. The improvement according to claim 4, comprising the additional steps of:delivering the transportation receptacle to a drug utilization facility;at said facility, removing the outer container from the transportation receptacle;delivering the removed outer container to a drug utilization area within the facility; andwithin the drug utilization area, sterilizing the protective outer container before removing the pig from the container.
	summary
	summary
	claims	1. An X-ray exposure method comprising: supporting an X-ray mask unit in which a patterned X-ray absorber is formed on a membrane, said patterned X-ray absorber containing one of an element having a density/atomic weight of not less than 0.085 [g/cm 3 ] and an L-shell absorption edge at a wavelength of 0.75 to 1.6 nm and an element having a density/atomic weight of not less than 0.04 [g/cm 3 ] and an M-shell absorption edge at a wavelength of 0.75 to 1.6 nm; and applying synchrotron radiation having maximum light intensity at a wavelength of 0.6 to 1 nm onto said X-ray mask unit. 2. A method according to claim 1 , wherein said patterned X-ray absorber is a material containing at least one element selected from the group of Co, Ni, Cu, Zn, Ga, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, and Lu. claim 1 3. A method according to claim 1 , wherein said patterned X-ray absorber is a material containing at least one element selected from the group of Cu, Ni, and Zn. claim 1 4. A method according to claim 1 , wherein a patterned transparent film different from the patterned X-ray absorber is further formed on said membrane. claim 1 5. A method according to claim 4 , wherein said patterned transparent film is a material containing at least one selected from the group of SiO 2 , SiC, Si, SrO, and SiON. claim 4 6. An X-ray exposure method comprising: supporting an X-ray mask unit in which a patterned X-ray absorber is formed on a membrane, said patterned X-ray absorber being formed of one of an alloy and a multi-layer film, which comprises a first material containing an element having an L-shell absorption edge or an M-shell absorption edge at a wavelength of 0.75 to 1.6 nm and a second material containing an element having an M-shell absorption edge at a wavelength of 0.5 to 0.75 nm; and applying synchrotron radiation having maximum light intensity at a wavelength of 0.6 to 1 nm onto said X-ray mask unit. 7. A method according to claim 6 , wherein said first material contains at least one element selected from the group of lanthanoid rare-earth elements of atomic numbers 57 to 71 and, Co, Ni, Cu, Zn, and Ga, and said second material contains at least one element selected from the group of Hf, Ta, W, Re, Os, Ir, Pt, Au, and Hg of atomic numbers 72 to 80. claim 6 8. A method according to claim 6 , wherein a patterned transparent film different from the patterned X-ray absorber is further formed on said membrane. claim 6 9. A method according to claim 8 , wherein said patterned transparent film is a material containing at least one element selected from the group of SiO 2 , SiC, Si, SrO, and SION. claim 8 10. An X-ray exposure method comprising: supporting an X-ray mask unit in which a patterned X-ray absorber is formed on a membrane, said patterned X-ray absorber being a material containing as a major constituent an element having all L- and M-shell absorption edges in a region shorter than the shortest wavelength or longer than the longest wavelength of an exposure wavelength region having an intensity not less than {fraction (1/10)} the light intensity at a wavelength of maximum light intensity of synchrotron radiation to be incident; and applying the synchrotron radiation onto said X-ray mask unit. 11. A method according to claim 10 , wherein the synchrotron radiation has maximum light intensity at a wavelength of 0.6 to 1 nm, and all the L- and M-shell absorption edges of the element exist in a region of not more than 0.65 nm and not less than 1.02 nm. claim 10 12. A method according to claim 11 , wherein said patterned X-ray absorber is a material containing as a major constituent at least one element selected from the group of Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, La, Ce, Pr, Nd, Pm, Sm, Eu, and Gd. claim 11 13. A method according to claim 11 , wherein said patterned X-ray absorber is a material containing as a major constituent at least one element selected from the group of Cu, Ni, and Zn. claim 11 14. A method according to claim 10 , wherein a patterned transparent film different from the patterned X-ray absorber is further formed on said membrane. claim 10 15. A method according to claim 14 , wherein said patterned transparent film is a material containing at least one element selected from the group of SiO 2 , SiC, Si, SrO, and SiON. claim 14 16. A method according to claim 14 , wherein a material, which makes the ratio of a deviation of maximum an minimum phase shift difference between the absorber and the transparent material with arbitrary thickness for the wavelength of the exposure wavelength from an average phase shift difference in the exposure wavelength region smaller than that of a deviation of maximum and minimum phase shift of the absorber with arbitrary thickness for the wavelength of the exposure wavelength from an average phase shift difference in the exposure wavelength region, is used as the material of said transparent film. claim 14
	claims	1. A method of smoothing a solid surface with a gas cluster ion beam comprising:an irradiation step of directing the gas cluster ion beam onto the solid surface at an irradiation angle not smaller than a critical angle, where the irradiation angle, is between a normal to the solid surface and the gas cluster ion beam;whereinthe critical angle is an irradiation angle of 70 degrees at which a distance of interaction between the solid and a cluster colliding with the solid turns to an abrupt increase; andthe irradiating step comprises a process of directing the gas cluster ion beam while varying the irradiation angle in a range not smaller than the critical angle. 2. The method of smoothing a solid surface with a gas cluster ion beam according to claim 1, wherein the irradiation step comprises:a process of directing the gas cluster ion beam while varying a dose of the gas cluster ion beam. 3. The method of smoothing a solid surface with a gas cluster ion beam according to claim 1, wherein the irradiation step comprises a process of directing the gas cluster ion beam of a dose determined with reference to a database that allows a dose to be determined from at least a desired etching amount and the irradiation angle. 4. The method of smoothing a solid surface with a gas cluster ion beam according to claim 1, wherein, when a striped uneven pattern is present on the solid surface, the irradiation step comprises a process of directing the gas cluster ion beam in a direction nearly parallel to the direction in which the uneven pattern is repeatedly formed. 5. A solid surface smoothing apparatus that uses a gas cluster ion beam comprising:gas cluster ion beam emission means adapted to emit the gas cluster ion ea onto the solid surface; andan irradiation angle setting means adapted to set an irradiation angle to a critical angle or a greater angle, the irradiation angle being between a normal to the solid surface and the gas cluster ion beam,whereinthe critical angle is an irradiation angle of 70 degrees at which a distance of interaction between the solid and a cluster colliding with the solid turns to an abrupt increase; andthe irradiation angle setting means is adapted to vary the irradiation angle in a range not smaller than the critical angle. 6. The solid surface smoothing apparatus according to claim 5, further comprising:a database that allows a dose to be determined from at least a desired etching amount and the irradiation angle;wherein the gas cluster ion beam emission means emits the gas cluster ion beam of the dose determined with reference to the database in accordance with the desired etching amount and the irradiation angle specified by the irradiation angle setting means. 7. A method of smoothing a solid surface with a gas cluster ion beam comprising:an irradiation step of directing the gas cluster ion beam onto the solid surface at an irradiation angle not smaller than a critical angle, where the irradiation angle is between a normal to the solid surface and the gas cluster ion beam,whereinthe critical angle is an irradiation angle at which a distance of interaction between the solid and a cluster colliding with the solid turns to an abrupt increase; andthe irradiation step comprises, when a striped uneven pattern with interval within a range of 800 nm to 10 μm inclusive is present on the solid surface, a process of directing the gas cluster ion beam in a direction nearly parallel to the direction in which the uneven pattern is repeatedly formed. 8. The method of smoothing a solid surface with a gas cluster ion beam according to claim 7, wherein the critical angle is 70 degrees.
	description	The present invention relates to a transport container for a non-irradiated nuclear fuel assembly. New (or non-irradiated) nuclear fuel assemblies are generally manufactured at a production site and then transported to a nuclear power station. During transport, the nuclear fuel assemblies have to be protected in order to preserve their integrity under normal transport conditions, for the purpose of ensuring their later use in a nuclear reactor under the required conditions of safety and performance, and in order to minimize the risk of malfunction in the event of an accident in the course of transport, in particular in order to avoid dispersing fissile material and approaching conditions of criticality. FR 2 774 800 describes a transport container for nuclear fuel assemblies comprising internal packaging delimiting two individual housings for nuclear fuel assemblies, and external packaging formed from two half-shells, the internal packaging being suspended inside the external packaging. The object of that arrangement is to protect new nuclear fuel assemblies from impact and vibration thanks to the two nested packagings and to the means of suspension between the two packagings. Nevertheless, that container is particularly bulky and heavy. It can be stored only horizontally on the ground and the large surface area which it occupies on the ground limits the number of containers which can be stored intermediately at the site of a nuclear power station. Its large dimensions force operators to work high above the ground, for example in order to lash down the container on a transport platform. Finally, a large number of journeys is necessary to ensure the delivery of the fuel in compliance with regulations. Those characteristics increase the cost of transporting and exploiting that type of container considerably. An object of the present invention is to provide a transport container for a nuclear fuel assembly which enables the transport costs to be reduced considerably and the conditions of exploitation thereof to be improved. To that end, the invention provides a transport container for a non-irradiated nuclear fuel assembly, characterized in that it comprises a single casing for receiving at least one nuclear fuel assembly, the casing being formed from an elongate tubular shell, the shell comprising a metallic internal layer delimiting at least one individual housing for receiving a nuclear fuel assembly, and a metallic external layer surrounding the internal layer, the shell being filled between its internal layer and its external layer, and from lids for closing the or each housing at the longitudinal ends of the shell. According to other embodiments, the container comprises one or more of the following features, taken individually or in accordance with any technically possible combination: it comprises lateral faces which delimit the or each individual housing and which are substantially smooth over the entire length of the housing; the transverse clearance between lateral faces of the or each individual housing and an assembly which is to be received inside the housing is selected in such a manner that damage to the assembly caused by relative movement in the housing is avoided; the transverse clearance between the lateral faces of the or each individual housing and an assembly which is to be received inside the housing is from 0.1 to 5 mm, preferably from 0.3 to 3 mm, and even more preferably from 0.5 to 1 mm; the transverse clearance is adjusted by adding plates which are secured to the internal surfaces of the housings; the or each housing has a quadrangular cross-section defined by two pairs of lateral faces arranged in the shape of a V, the shell comprising for the or each housing a support carrying one of the pairs of lateral faces arranged in the shape of a V, a door carrying the other of the two pairs of lateral faces arranged in the shape of a V, and means for securing the door to the support enabling the position of the door and the support to be adjusted in the transverse direction passing via the lines of intersection of the pairs of faces arranged in the shape of a V; the shell comprises an intermediate layer for neutrophage insulation and an intermediate layer for protection against impact; a layer for thermal protection is added to the layer for protection against impact; the protective layer surrounds the neutrophage insulation layer; and the shell comprises an intermediate metallic separation layer separating the neutrophage insulation layer and the protective layer. The invention also provides a method of transporting at least one nuclear fuel assembly, wherein at least one non-irradiated nuclear fuel assembly is placed in a container such as defined above, the container is placed on a transport vehicle in such a manner that the casing of the container protects by itself the nuclear fuel assembly in the event of the container falling, and the container is transported from a first site to a second site. According to one embodiment, suspension members are arranged between the shell and a deposit surface. The nuclear fuel assembly 2 of FIG. 1 is of the type which is to be used in pressurized light water nuclear reactors (PWR). The assembly 2 is elongate in a longitudinal direction L. It comprises a bundle of nuclear fuel rods 4, and a framework 5 for supporting the rods 4. The assembly 2 has a square cross-section in the example illustrated. The rods 4 are in the form of tubes which are filled with nuclear fuel pellets and which are closed at their ends by plugs. As is conventional, the framework 5 comprises two end-pieces 6 arranged at the longitudinal ends of the assembly 2, guide tubes (not shown) extending longitudinally between the end-pieces 6, and grids 8 for holding the rods 6. The guide tubes are secured at their ends to the end-pieces 6. The grids 8 are secured to the guide tubes and distributed between the end-pieces 6. The rods 4 extend through the grids 8 which hold them longitudinally and transversely. The nuclear fuel assemblies are generally manufactured at a manufacturing site and then transported to a nuclear power station where they are stored intermediately before being introduced into the core of a nuclear reactor. FIG. 2 illustrates a container 10 according to the invention enabling the assembly 2 to be transported from a manufacturing site to a nuclear power station. The container 10 comprises a single casing 11 for receiving nuclear fuel assemblies, the casing 11 being formed from a shell 12 and two lids 13. The shell 12 is tubular and longitudinally elongate in a longitudinal direction E. It has an internal surface 14 delimiting an internal cavity, and an external surface 16. The shell 12 comprises a longitudinal partition 18 separating the internal cavity of the shell 12 into two distinct and separate individual housings 20. The housings 20 extend parallel with each other one on each side of the partition 18, in the longitudinal direction E of the shell 12. Each housing 20 is to receive a nuclear fuel assembly 2 such as that of FIG. 1, and has a corresponding cross-section, here a square cross-section. Each housing 20 can, if necessary, receive a rod case containing fuel rods which are to be assembled to form fuel assemblies, or which are to be used to repair previously delivered fuel assemblies. The lids 13 are provided to close the housings 20 at the longitudinal ends of the shell 12. As is conventional, each lid 13 comprises metallic plates enclosing an energy-dissipating material, such as balsa wood, a foam or a honeycomb structure. The lids 13 form devices for dissipating energy in the event of the container 10 falling axially. By way of variation, they have an internal structure similar to that of the shell 12, which is described hereinafter. The internal structure of the shell 12 is illustrated in FIGS. 3 and 4 which are cross-sectional views of the shell 12. As shown in FIG. 3, the shell 12 is layered and comprises several superposed layers. To be more precise, it comprises, from the inside to the outside of the shell 12, a metallic internal layer 22, a neutron insulating layer 24, a metallic intermediate layer 26, a protective layer 28 and a metallic external layer 30. The intermediate layers 24, 26 and 28 fill the shell 12 between the internal layer 22 and the external layer 30, substantially without leaving a gap. The internal layer 22 defines the housings 20. The external layer 30 defines the external skin of the shell 12. The intermediate layer 26 separates the insulating layer 24 and the protective layer 28. The protective layer 28 is to absorb the energy of violent mechanical impact and to insulate the fuel assemblies 2 thermally with respect to the outside. It is preferably solid in order to contribute to the mechanical strength of the shell 12. It is formed, for example, by a high-density foam or by balsa wood. The insulating layer 24 is a neutrophage layer which is to absorb neutrons emitted by nuclear fuel assemblies received in the housings 20. It is preferably solid in order to contribute to the mechanical strength of the shell 12. The insulating layer 24 is, for example, a resin which is charged with a neutrophage chemical compound or element, such as boron, and which is rich in hydrogen, or is a sheet of neutron-absorbing metallic material, such as hafnium. In a variant of the invention, the layers 22, 24 and 26 are replaced by a single thick metallic internal layer composed, for example, of boron steel or boron aluminium (a few % boron by mass). The shell 12 optionally comprises reinforcing members 31, here in the form of longitudinal beams arranged in the thickness of the shell 12, between the internal layer 22 and the external layer 30. The reinforcing members 31 are, for example, metal profiles having an “I”-shaped cross-section, one foot of which is secured to the internal layer 22 and the core of which extends radially towards the outside. Other suitable means of mechanical reinforcement may be envisaged. Thus, by way of variation or optionally, internal reinforcing members are provided in the form of flat profiles, tubes, angled members or corrugated metal sheets which are each placed between two metallic layers, for example between the metallic intermediate layer 26 and the metallic external layer 30 and/or between the metallic internal layer 22 and the metallic intermediate layer 26, and fixedly joined, for example by welding, thereto. By way of variation or optionally, the shell 12 comprises members for reinforcing the external layer, for example in the form of ribs projecting from the external layer towards the outside of the shell 12. The ribs are integral with the external layer 30 or are attached thereto. The shell 12 is formed from several shell portions which are elongate in the longitudinal direction E. To be more precise, the shell 12 is formed by a first shell portion forming a support 32 having a cross-section in the shape of a “T” and a second shell portion formed by two doors 34 having a cross-section in the shape of an “L”. The support 32 comprises the partition 18 defining the down-stroke of the “T” and two wings 38 extending symmetrically one on each side of the partition 18 and defining the cross stroke of the “T”. The partition 18 is formed by a rib of a portion of the internal layer 22 filled with the material of the insulating layer 24 of the wings 38. As a result, the housings 20 are separated by a neutron insulating layer, which prevents the initiation of a nuclear reaction between two nuclear fuel assemblies 2 located in the housings 20. Each door 34 is secured to one end of the partition 18 and to one end of a wing 38. Each housing 20 is defined by two faces 40 of the support 32 forming a V at 90°, and two faces 42 of one of the doors 34, forming a V at 90°. One of the faces 40 of the support 32 is a face of the partition 18, and the other a face of a wing 38. The faces 40 and 42 are smooth over the entire length of the housing 20. They define a cross-section corresponding substantially to that of the end-pieces 6 and the grids 8 of the assembly 2 of FIG. 1 which are to be received in the housings 20. The total lateral clearance between the faces 40, 42 and the grids 8 is from 0.1 to 5 mm, preferably from 0.3 to 3 mm, even more preferably from 0.5 to 1 mm, in order to avoid relative displacement of the assembly 2 in the housing 20, the amplitude of which would damage the assembly received in the housing, while at the same time permitting the longitudinal insertion of the assembly 2 into the housing 20. The container 10 is free from means for the transverse clamping of the assemblies 2 inside the housings 20. The small clearance existing between the assembly 2 and the housing 20 enables the amplitude of the relative movement between the assembly 2 and the housing 20 to be limited and the integrity of the assembly 2 to be preserved. In addition, it is possible to provide for a slight clamping of the assemblies 2 inside the housings 20 owing to the cooperation between the support 32 and the doors 34. To that end, as shown in FIG. 4 which illustrates a cross-sectional view in a plane different from that of FIG. 3, the shell 12 comprises means for securing each door 34 to the support 32 by clamping in a direction D substantially parallel with the a straight line L passing via the vertex of the V defined by the faces 42 of the door 34 and the vertex of the V defined by the faces 40 of the support 32. The securing members comprise, for example, screws 44 shown diagrammatically in FIG. 4. Resilient supports, for example of rubber, may be provided in the areas of contact between the doors 34 and the support 32. The external layer 30 comprises wells 46 extending from the external surface 16 as far as the internal layer 22, for the passage of the screws 44. The wells 46 advantageously form reinforcing stays extending between the external layer 30 and the internal layer 22. As schematically shown for example in FIG. 11, it is possible to adjust the size of the housing 20 by adding schematically shown adjusting plates 50 of suitable thickness to one or more of the lateral faces 40, 42 of the housings 20 in order to permit the transport of fuel assemblies of different sizes. Optionally, the upper lid 13 comprises a device for the longitudinal clamping of the assembly in order to avoid any displacement of the assembly in the longitudinal direction L of the assembly 2 in the housing 20 after positioning the upper lid 13 and closing the container 10 and during all handling and transport operations. As shown in FIG. 5 which illustrates an end view of the shell 12, the loading of assemblies 2 can be effected by placing the support 32 on the ground, disengaging the doors 34, placing the assemblies 2 on the support 32, then reclosing the doors 34 and clamping them on the support 32. The lids 13 may or may not be withdrawn, depending on the individual case, in order to effect this loading. Unloading can be carried out in the same manner. The small surface area occupied on the ground and the small bulk of the container facilitates the exploitation thereof. This variant of loading and unloading while the container 10 is in the horizontal position will, however, for the most part be used for rod cases. As shown in FIG. 6 which represents a perspective view, the loading or unloading of assemblies 2 can be effected by placing the container 10 in the vertical position, the support 32 being arranged against a wall or a support structure, the upper lid 13 being withdrawn, by gripping the assembly 2 by its upper end-piece 6, in known manner with appropriate lifting grippers, and by displacing the assembly 2 vertically in one of the housings 20. This is possible owing to the fact that the lateral faces 40, 42 are substantially smooth over the entire length and there is therefore no risk of the assembly 2 catching on a raised portion inside the housing 20. This method of loading or unloading permits a major space saving because it avoids storing the container 10 in a horizontal position, and a major time saving because it avoids removing the doors 34: only the upper lid 13 has to be removed. In a variant illustrated in FIG. 7, the doors 34 are articulated on the wings 38 of the support 32 by way of hinges 150 having longitudinal axes. It is possible to load such a container in the manner illustrated in FIG. 6, from the top, or in the manner illustrated in FIG. 8, from the side. In order to do this, with the container 10 being placed in a vertical position and the upper lid 13 being withdrawn, a door 34 is opened to insert or withdraw the assembly 2. That method of loading is more suitable when the height of the building is limited. In order to enable a door 34 to be opened when the container 10 is in the vertical position, it is provided that the lower lid 13 of the container 10 is not connected to the door 34 during the loading or unloading operation. As shown in FIG. 9, it is possible to provide in a nuclear power station an intermediate storage device of the rack type, enabling a plurality of containers 10 to be stored intermediately in an upright position next to each other, with a particularly large space saving compared with storage in a horizontal position, and without a time limit. For, with a conventional container permitting storage only in the horizontal position, the duration of storage is limited in order to avoid damage to the assembly which is not designed to be stored horizontally. The containers must ensure that, in the course of transport, the nuclear fuel assemblies are protected against impact, especially when accidents occur. The fuel assembly transport containers must pass very strict qualification tests defined by international standards, such as the Regulations for the transport of radioactive material of the International Atomic Energy Agency (IAEA), the European Agreement concerning the international carriage of Dangerous goods by Road (ADR), the Regulations concerning the international carriage of dangerous goods by rail (RID), and the International Maritime Dangerous Goods (IMDG) Code. In particular, the containers undergo especially a regulation drop test from a height of 9 m, with in general a first drop, the longitudinal direction of the container being in a vertical position, and a second drop, the longitudinal direction of the container being inclined relative to the horizontal by a predetermined value in order to cause the maximum damage and the inclination value being, for example, of the order of 15° in order to obtain a whipping effect, and a regulation drop test from a height of 1.5 m onto a punch, in the course of which tests the internal layer 22 must not open. The casing 11 formed by the shell 12 and the lids 13 is suitable by itself for transporting non-irradiated nuclear fuel assemblies. In other words, it is capable by itself of protecting the nuclear fuel assemblies it contains in the course of transport, and in particular in the course of the regulation drop tests. For, the multi-layered structure of the shell 12 confers on it a significant mechanical impact strength. The mechanical strength of the shell 12 is determined, among other things, by the following dimensions: thickness of the metal sheets forming the metallic layers 22, 26, 30; material and thickness of the insulating layer 24; material and thickness of the protective layer 28; material, shape, dimensions, number and position of the reinforcing members 31. The thicknesses of the various layers and the nature thereof can be easily determined by the person skilled in the art on the basis of the characteristics of the fuel assemblies to be transported. The nature and thickness of the insulating layer 24 will depend, for example, on the uranium 235 enrichment of the fissile material contained in the fuel rods 4. A criticality study will easily enable the most appropriate material and the necessary thickness to be determined: a sheet of hafnium a few hundredths to a few tenths of a mm thick or a sheet of boron aluminium or steel with a few percent of boron and a thickness of a few millimeters (for example from 1.5 to 4 mm) associated with a few centimeters of moderator material, such as polyethylene, equivalent to several centimeters (50 to 75 mm) of resin charged with a few percent of boron. The thickness of the metal sheets 22, 26, 30 forming the metallic layers, and the material, shape, dimensions, number and position of the reinforcing members 31 will be determined, for example, by means of strength calculations in respect of the materials in order to avoid any deformation of the container and the transported assemblies and to ensure the necessary mechanical strength to pass the regulation tests successfully. By way of example, the thicknesses of the metal sheets forming the metallic layers 22, 26, 30 and of any reinforcing members 31 provided in the form of bent metal sheets will advantageously be from 1 to 6 mm. Reinforcing members 31 in the form of an “I”-shaped beam may be, for example, in the range 80-140 according to Euronorm 19-57. The protective layer 28 arranged on the outside protects against impact and prevents the insulating layer 24 from being affected and deformed excessively in the event of impact. Its thickness will advantageously be from 30 to 150 mm. The insulating layer 24 absorbs the neutrons emitted by the fissile material and prevents them from being scattered outside the container 10. All of the layers are grouped together as close as possible to the assemblies 2 received in the housings 20, without a gap between those layers. This makes it possible to limit the bulk and the mass of the container 10 by reducing the volume of material, to increase the resistance of the shell 12 to mechanical impact, and to limit the energy to be dissipated in the event of the container 10 falling. The containers must also protect the nuclear assemblies in the event of a fire under normal transport conditions (undamaged container) but also under transport conditions after the occurrence of an accident (container damaged after a fall in accordance with the above-mentioned regulation tests). In particular, the containers undergo a regulation fire resistance test in the course of which they have to withstand for 30 minutes a temperature of 850° C. brought about by a hydrocarbon fire. The shell 12 and the lids 13 are capable by themselves of protecting nuclear fuel assemblies in the event of a fire. The layered structure of the shell 12 ensures effective thermal protection in order to avoid an increase in temperature which could damage the assembly and its components. The protective layer 28 is also configured to form a barrier against the propagation of heat from the outside to the inside of the shell 12. In a variant, a specific thermal insulation layer is added between the metallic layers 26 and 30. FIG. 12 schematically shows a layer for thermal protection 48 added to the second intermediate layer 28. Thus, the shell 12 and the lids 13 are capable by themselves of protecting the non-irradiated nuclear fuel assemblies in the various regulation tests to which the containers are subjected before being qualified for the road, rail, marine or air transport of non-irradiated nuclear fuel assemblies. It is not necessary to provide overpackaging. The container 10 is particularly compact and light. As a result, its manipulation facilitated, as is its transport. A larger number of containers 10 can be placed on the same transport means, such as a lorry, a railway wagon or a marine or air container. The costs of transport and exploitation are therefore reduced. The amount of neutron absorber and its presence as close as possible to the fuel assemblies also prevents any risk of starting a nuclear reaction between several containers 10 loaded, transported or stored together, without any limitation as to numbers. As shown in FIG. 2, the container 10 comprises securing members for its manipulation, its lashing-down and its transport. The container 10 comprises two tubular feet 52 secured transversely via a first face 16A of the external surface 16 of the container 10 to a reinforcing member 31. The feet 52 are configured to permit the engagement and the locking of securing members installed on the transport platform concerned (lorry, wagon, marine or air container) or on the container or the intermediate structure arranged below the container 10. The container 10 comprises securing members 54 secured to a reinforcing member 31 via a second face 16B of the external container surface 16 opposite the first face. Those securing members 54 are to be secured to the feet of another container stacked on the container 10 or on an intermediate structure. The container 10 comprises on the second face 16B, tubes 56 for receiving the forks of a lifting truck in order to enable the container to be lifted and placed on a lorry or a wagon. Those tubes 56 are arranged to receive handling tool securing members and also to permit the handling of the container by a suitable lifting means (rolling bridge, crane) and the vertical loading/unloading of the transport platform. As illustrated in FIG. 10, according to one method of transport, at least one non-irradiated nuclear fuel assembly 2 is placed in the container 10, the container 10 is placed on a vehicle 58, in particular a road transport vehicle, and the container 10 is transported from a first site (for example a manufacturing site) to a second site (for example a nuclear power station) using, if necessary, intermodal means (road, rail, marine and/or air transport). The container 10 is placed on the vehicle 58 in such a manner that the shell 12 protects the assemblies 2 by itself in the event of the container falling. Thus, no closed additional packaging and no overpackaging are arranged around the shell 12. The impact strength of the container 10, which is conferred on it by the multi-layered structure of its shell 12, further enhanced by the protective layer 28, and the neutron insulation which is conferred on it by its insulating layer 24, enable it to be transported without overpackaging. Furthermore, the assemblies have to be protected from vibration which could impair the support of the fuel rods 4 in the holding grids 8 or which could cause an axial displacement of the fuel pellet column and prevent or impair the later use of the assembly in a nuclear reactor. The integrity of the nuclear fuel assembly can be preserved efficiently with a container 10 according to the invention, which comprises an internal layer 22 connected rigidly to an external layer 30, without suspension members, and in which each assembly 2 is held in a corresponding housing 20 purely owing to the small clearance between the assembly and the faces 40, 42 delimiting the housing 20. It is possible to provide suspension members 60 between the transport platform 62 of the vehicle 58 and the container 10 in order to filter the vibration caused by transport. The suspension members 60 are, for example, simple elastomeric blocks. It is also possible to provide suspension members between the container 10 and any deposit surface, whether this be another container or an intermediate structure located below the container 10 when the containers are stacked on the transport platform 62.
	description	This invention was made with Government support under contract number DE-FC52-09NA29626, awarded by the U.S. Department of Energy. The Government has certain rights in the invention. Elements, and specific isotopes thereof, may be formed by bombarding parent materials with appropriate radiation to cause a conversion to desired daughter isotopes. For example, precious metals and/or radioisotopes may be formed through such bombardment. Conventionally, particle accelerators or specially-designed, non-commercial test reactors are used to achieve such bombardment and produce desired isotopes in relatively small amounts. Radioisotopes have a variety of medical and industrial applications stemming from their ability to emit discreet amounts and types of ionizing radiation and form useful daughter products. For example, radioisotopes are useful in cancer-related therapy, medical imaging and labeling technology, cancer and other disease diagnosis, and medical sterilization. Radioisotopes having half-lives on the order of days or hours are conventionally produced by bombarding stable parent isotopes in accelerators or low-power, non-electricity-generating reactors. These accelerators or reactors are on-site at medical or industrial facilities or at nearby production facilities. Especially short-lived radioisotopes must be quickly transported due to the relatively quick decay time and the exact amounts of radioisotopes needed in particular applications. Further, on-site production of radioisotopes generally requires cumbersome and expensive irradiation and extraction equipment, which may be cost-, space-, and/or safety-prohibitive at end-use facilities. Example embodiments include systems for delivering and retrieving irradiation targets in otherwise inaccessible nuclear reactors to generate desired isotopes from the irradiation targets. Example systems include a penetration pathway from an origin to an instrumentation tube in the reactor to permit irradiation targets to move between the two, and a loading/offloading system that can select between loading and offloading paths to guide fresh and irradiated targets to/from different destinations and locations based on their status. The origin and harvesting destination may be outside of an access-restricted area to permit access for refilling or harvesting during plant operation. As an example, penetration pathways and loading/offloading systems may include tubing that provides an enclosed and thus air-tight route for irradiation targets to traverse, with filtered exhaust points that prevent migration of radioactive progeny out of example systems. Example systems may be gravity-driven and/or may further be useable with, or include, a drive system that can actively move irradiation targets between destinations. Loading/offloading systems may accommodate drive systems by being attachable thereto and providing additional paths for drive systems to enter into penetration pathways. For example, a drive system may include a plunger and a cable shaped to pass through the penetration pathway and mechanically push against a discreet irradiation target, a pneumatic system, a conveyor, a magnetic system, etc. and may take advantage of existing Traversing Incore Probe (TIP) drives. Loading/offloading systems may provide several different paths to distinct destinations through the use of a loading junction, T-junction, turntable, Y-junction, selector, etc. Flow restrictors and sensors throughout example systems may enable automatic path switching and driving to move irradiation targets through example systems from source to irradiation tube to harvesting area. Similarly, irradiation targets may be provided manually or automatically from a reservoir based on need and plant operation. Harvesting areas may include a DoT and/or USNRC-compliant shipping cask for containing and storing irradiated irradiation targets containing desired daughter products for end-user consumption. Irradiation targets may be shaped, sized, and otherwise configured to be compatible with movement through example systems. For example, irradiation targets may be formed of spheres of a solid material that will convert to a useable amount of daughter product when exposed to neutron radiation in an instrumentation tube for a number of hours or days. Molybdenum-98, natural iridium, or liquid or gaseous materials, for example, may be used as irradiation targets given proper containment. Irradiation targets are further configured to fit within instrumentation tubes and may be held in the same in example systems by drive systems and/or static latches, valves, gravity, pneumatic pressure, etc. Example methods include creating a penetration pathway to an instrumentation tube for traverse by an irradiation target, moving the irradiation target into the instrumentation tube through the penetration pathway, irradiating the irradiation target in the instrumentation tube, switching to an exit pathway between the instrumentation tube and a harvesting area outside of an access barrier, and moving the irradiated irradiation targets to the harvesting area through the exit pathway for harvesting. This is a patent document, and general broad rules of construction should be applied when reading and understanding it. Everything described and shown in this document is an example of subject matter falling within the scope of the appended claims. Any specific structural and functional details disclosed herein are merely for purposes of describing how to make and use example embodiments. Several different embodiments not specifically disclosed herein fall within the claim scope; as such, the claims may be embodied in many alternate forms and should not be construed as limited to only example embodiments set forth herein. It will be understood that, although the terms first, second, etc. may be used herein to describe various elements, these elements should not be limited by these terms. These terms are only used to distinguish one element from another. For example, a first element could be termed a second element, and, similarly, a second element could be termed a first element, without departing from the scope of example embodiments. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. It will be understood that when an element is referred to as being “connected,” “coupled,” “mated,” “attached,” or “fixed” to another element, it can be directly connected or coupled to the other element or intervening elements may be present. In contrast, when an element is referred to as being “directly connected” or “directly coupled” to another element, there are no intervening elements present. Other words used to describe the relationship between elements should be interpreted in a like fashion (e.g., “between” versus “directly between”, “adjacent” versus “directly adjacent”, etc.). Similarly, a term such as “communicatively connected” includes all variations of information exchange routes between two devices, including intermediary devices, networks, etc., connected wirelessly or not. As used herein, the singular forms “a”, “an” and “the” are intended to include both the singular and plural forms, unless the language explicitly indicates otherwise with words like “only,” “single,” and/or “one.” It will be further understood that the terms “comprises”, “comprising,”, “includes” and/or “including”, when used herein, specify the presence of stated features, steps, operations, elements, ideas, and/or components, but do not themselves preclude the presence or addition of one or more other features, steps, operations, elements, components, ideas, and/or groups thereof. It should also be noted that the structures and operations discussed below may occur out of the order described and/or noted in the figures. For example, two operations and/or figures shown in succession may in fact be executed concurrently or may sometimes be executed in the reverse order, depending upon the functionality/acts involved. Similarly, individual operations within example methods described below may be executed repetitively, individually or sequentially, so as to provide looping or other series of operations aside from the single operations described below. It should be presumed that any embodiment having features and functionality described below, in any workable combination, falls within the scope of example embodiments. FIG. 1 is an illustration of a conventional nuclear reactor pressure vessel 10 usable with example embodiments and example methods. Reactor pressure vessel 10 may be, for example, a 100+ MWe commercial light water nuclear reactor conventionally used for electricity generation throughout the world. Reactor pressure vessel 10 is conventionally contained within an access barrier 411 that serves to contain radioactivity in the case of an accident and prevent access to reactor 10 during operation of the reactor 10. As defined herein, an access barrier is any structure that prevents human access to an area during operation of the nuclear reactor due to safety or operational hazards such as radiation. As such, access barrier 411 may be a containment building sealed and inaccessible during reactor operation, a drywell wall surrounding an area around the reactor, a reactor shield wall, a human movement barrier preventing access to instrumentation tube 50, etc. A cavity below the reactor vessel 10, known as a drywell 20, serves to house equipment servicing the vessel such as pumps, drains, instrumentation tubes, and/or control rod drives. As shown in FIG. 1 and as defined herein, at least one instrumentation tube 50 extends into the vessel 10 and near, into, or through core 15 containing nuclear fuel and relatively high levels of neutron flux and other radiation during operation of the core 15. As existing in conventional nuclear power reactors and as defined herein, instrumentation tubes 50 are enclosed within vessel 10 and open outside of vessel 10, permitting spatial access to positions proximate to core 15 from outside vessel 10 while still being physically separated from innards of the reactor and core by instrumentation tube 50. Instrumentation tubes 50 may be generally cylindrical and may widen with height of the vessel 10; however, other instrumentation tube geometries may be encountered in the industry. An instrumentation tube 50 may have an inner diameter of about 1-0.5 inch, for example. Instrumentation tubes 50 may terminate below the reactor vessel 10 in the drywell 20. Conventionally, instrumentation tubes 50 may permit neutron detectors, and other types of detectors, to be inserted therein through an opening at a lower end in the drywell 20. These detectors may extend up through instrumentation tubes 50 to monitor conditions in the core 15. Examples of conventional monitor types include wide range detectors (WRNM), source range monitors (SRM), intermediate range monitors (IRM), and traversing Incore probes (TIP). Access to the instrumentation tubes 50 and any monitoring devices inserted therein is conventionally restricted to operational outages due to containment and radiation hazards. Although vessel 10 is illustrated with components commonly found in a commercial Boiling Water Reactor, example embodiments and methods are useable with several different types of reactors having instrumentation tubes 50 or other access tubes that extend into the reactor. For example, Pressurized Water Reactors, Heavy-Water Reactors, Graphite-Moderated Reactors, etc. having a power rating from below 100 Megawatts-electric to several Gigawatts-electric and having instrumentation tubes at several different positions from those shown in FIG. 1 may be useable with example embodiments and methods. As such, instrumentation tubes useable in example methods may be at any geometry about the core that allows enclosed access to the flux of the nuclear core of various types of reactors. Applicants have recognized that instrumentation tubes 50 may be useable to relatively quickly and constantly generate short-term radioisotopes on a large-scale basis without interfering with an operating or refueling core 15. Applicants have further recognized a need to generate short-term radioisotopes and remove them from within access barrier 411 quickly, without having to shut down an operating nuclear reactor to access an area within access barrier 411. Example methods include inserting irradiation targets into instrumentation tubes 50 and exposing the irradiation targets to the core 15 while operating or producing radiation, thereby exposing the irradiation targets to the neutron flux and other radiation commonly encountered in the operating core 15. The core flux over time converts a substantial portion of the irradiation targets to a useful mass of radioisotope, including short-term radioisotopes useable in medical applications. Irradiation targets may then be withdrawn from the instrumentation tubes 50, even during ongoing operation of the core 15, and removed for medical and/or industrial use. Example embodiment systems include a penetration pathway and loading/offloading system to provide irradiation targets in the above manner from origin points and to terminal points accessible during plant operation. Drive systems and several other components are useable with example embodiment systems, and specific example embodiments and methods are discussed below, including details of how example methods may enable example embodiments. FIG. 2 is a schematic drawing of an example embodiment irradiation target delivery and retrieval system 1000 having specific types of a penetration pathway, a loading/offloading system, and a drive system. FIG. 2 illustrates various components of example system 1000 in a loading configuration, with other configurations possible and shown, in part, in other drawings. As shown in FIG. 2, example embodiment irradiation target delivery and retrieval system 1000 may include or use one or more elements to facilitate irradiation target loading, irradiation, and harvesting in a timely, automatic, and/or consumption-enhancing manner. System 1000 includes a penetration pathway that provides a path from outside access barrier 411 to instrumentation tube 50 for one or more irradiation targets, a loading/offloading system that permits new irradiation targets to be inserted and irradiated targets to be harvested outside access barrier 411, and a drive system that moves irradiation targets between instrumentation tube 50 and loading/offloading in example embodiment system 1000. A penetration pathway in example embodiment system 1000 provides a reliable path of travel for irradiation targets 250 between an accessible location, such as an offloading or loading area outside access barrier 411 into an instrumentation tube 50, so irradiation targets 250 can move within the pathway to a position in or near an operating nuclear core 15 for irradiation. Example pathways can include many delivery mechanisms used alone or in combination, including tubing, frames, wires, chains, conveyors, etc. in example embodiment system 1000 to provide a transit path for an irradiation target between an accessible location and an operating nuclear core. As a specific example shown in FIG. 2, a penetration pathway may include penetration tubing 1100 running between, either in portions or continuously, a loading junction 1200 and instrumentation tube 50 in a nuclear reactor. Penetration tubing 1100 may be flexible or rigid and sized to appropriately permit irradiation targets 250 to enter into and/or through penetration tubing 1100 and navigate various structures and penetrations in and within access barrier 411. Penetration tubing 1100 may be continuously sealed or include openings, such as at connecting junctions. Penetration tubing 1100 may junction with other tubes and/or structures and/or include interruptions. One possible advantage of penetration tubing 1100 being sealed and securely mating at junctures and/or with any terminal/originating points is that penetration tubing 1100 better maintains pneumatic pressure that can be used for target withdrawal, and also may provide additional containment for irradiation targets 250 and any products (gas, fluid, solid, particulate, etc.) formed as irradiation products in example embodiment system 1000. Penetration tubing 1100 may be fabricated of a material that maintains its physical characteristics in an operating nuclear reactor environment and does not significantly react with or entrain materials from irradiation targets 250 coming into contact therewith, including, for example, aluminum, stainless steel, carbon steel, nickel alloys, PVC, PFA, rubber, etc. Penetration tubing 1100 may be cylindrical or any other shape that permits irradiation targets 250 to enter into and/or pass through penetration tubing 1100. For example, penetration tubing 1100 may have a generally circular cross section with a 0.5-inch diameter and smooth interior surface that permits spherical irradiation target 250 to roll within penetration tubing 1100. One potential advantage of using such an example penetration tubing 1100 may be roughly matching diameters and geometries with instrumentation tube 50 for consistent irradiation target movement therein; however, alternate geometries, shapes, and sizes for penetration tubing 1100, or any other penetration pathway used in example embodiments, including those that limit movement, may be desirable, advantageous, and used. Penetration tubing 1100 used in example embodiment system 1000 provides a route from an origin at loading junction 1200, where irradiation targets may enter/exit penetration tubing 1100 outside of access barrier 411. As shown in FIG. 2, for example, penetration tubing 1100 leads irradiation targets 250 from loading junction 1200 to access barrier 411, which may be, for example, a steel-lined reinforced concrete containment wall or drywell wall or any other access restriction in conventional nuclear power stations. A penetration pathway flow limiter 1105, which may be a ball shear valve, a solenoid valve, a simple clamp, a pin and transducer, etc., may be placed on penetration tubing 1100 to control and/or prevent irradiation targets 250 from travelling in penetration tubing 1100 in one or more directions. For example, penetration pathway flow limiter 1105 may engage during a loading operation as shown in FIG. 2 to prevent irradiation targets from moving inside of access barrier 411 in penetration tubing 1100. Flow limiter 1105 may be positioned a distance L1 from any entry point of irradiation targets 250 from loading junction 1200 such that all irradiation targets are loaded into a penetration pathway before travelling therein. For example, L1 may be an equivalent distance in which irradiation targets 250 are intended to travel into instrumentation tube 50. Penetration pathway flow limiter 1105 may further seal any exterior pathway leading through access barrier 411 to reduce or prevent unwanted migration of material outside of access barrier 411 and provide isolation for areas within access barrier 411. Penetration pathways usable in example embodiment system 1000 provide a route through access barrier 411 and to reactor vessel 10 where irradiation targets 250 may enter an instrumentation tube 50. For example, as shown in FIG. 2, penetration tubing 1100 penetrates access barrier 411 and extends to instrumentation tube 50. Penetration tubing 1100 may pass through an existing penetration in access barrier 411, such as an existing TIP tube penetration, or may use a new penetration created for penetration tubing 1100. Penetration tubing 1100 negotiates or passes through any other objects inside of access barrier 411 before reaching instrumentation tube 50. An annular reactor pedestal 412 may be present in a drywell 20 beneath reactor 10, and penetration tubing 1100 is shown in FIG. 2 passing through a penetration in pedestal 412. It is understood that penetration pathways may follow any number of different courses and negotiate different obstacles in different reactor designs aside from the specific example path shown with penetration tubing 1100 in FIG. 2. Similarly, penetration pathways need not be consistent or uniform; for example, penetration tubing 1100 may terminate on either side of, and be connected to, a penetration in pedestal 412 to permit irradiation targets 250 to pass through the penetration between penetration tubing 1100. Penetration pathways useable in example embodiment system 1000 may terminate at or within an instrumentation tube. As shown in FIG. 2, penetration tubing 1100 terminates at a flange 1110 at a base of instrumentation tube 50, permitting irradiation targets 250 to pass from penetration tubing 1100 into instrumentation tube 50. Alternatively, penetration tubing 1100 may pass into and/or throughout instrumentation tube 50 to provide a liner or separate tube within instrumentation tube 50; in this way, penetration tubing 1100 may present a continuous inner diameter between a penetration pathway and instrumentation tube 50 and ensure that any variation or undesired geometry in instrumentation tube 50 does not significantly interfere with irradiation target 250 movement therein, while providing an additional level of containment for irradiation targets 250. As shown in FIG. 2, penetration tubing 1100 may descend a vertical distance H1 between flange 1110 of instrumentation tube 50 and a lowest point adjacent to access barrier 411. Distance H1 may have a length sufficient to ensure that all irradiation targets 250 exit instrumentation tube 50 by gravity and be at a point below flange 1110; that is, H1 may be greater than or equal to a maximum length of corresponding instrumentation tube 50. Configuring penetration tubing 1100 with a vertical height H1 may present an advantage of working with pneumatic driving system 500 and related components 510, 509, 501, and 502 by ensuring all irradiation targets 250 drop by gravity below flange 1110 and penetration valve 510 where a pneumatic driving fluid may be introduced. Penetration pathways useable in example embodiments may be pre-existing in part and/or installed during access to containment areas and/or restricted access areas in a nuclear power plant, such as during a pre-planned outage. For example, penetration tubing 1100 may be installed in access barrier 411 during an outage, with penetration tubing 1100 being passed through penetrations in access barrier 411 and pedestal 412, moved and secured in an area within access barrier 411 and a drywell space 20 under reactor 10, and secured to flange 1105. Portions of penetration tubing 1100 extending outside access barrier 411 may be installed at loading junction 1200 at any time. Penetration tubing 1100 may be secured at various points inside access barrier 411 and/or divert around existing equipment to minimize congestion or clutter in a drywell 20 or other space bounded by access barrier 411 while preserving a traversable path for irradiation targets 250 to and from instrumentation tube 50. Again, other penetration pathways, including wire guides, meshes, compartments, bored tunnels, etc. are useable in example embodiments to provide a path from outside an access-restricted area such as containment to an instrumentation tube of an operating nuclear reactor. Example embodiment irradiation target delivery and retrieval system 1000 further includes a loading/offloading system that permits new irradiation targets to be inserted and irradiated targets to be harvested outside access barrier 411. Loading and offloading systems usable with example embodiments permit irradiation targets 250 to be supplied and loaded into a penetration pathway and ultimately an instrumentation tube 50 from points outside of access barrier 411, as well as permitting irradiated irradiation targets 250 to be harvested outside of access barrier 411. As such, loading and offloading systems provide multiple paths between plural destinations in example embodiment systems and permit irradiation targets to be moved toward destinations based on path configuration and target irradiation status. Loading and offloading systems work with penetration pathways in example embodiments and may be connected to penetration pathways to deliver/receive irradiation targets to/from penetrations pathways. As shown in FIG. 2, an example loading and offloading system may include an irradiation target reservoir 1270 and reservoir flow limiter 1250. Reservoir 1270 may hold a particular or arbitrary number of irradiation targets 250 and be reloaded based on need or schedule with additional irradiation targets 250, regardless of plant operational status. Reservoir flow limiter 1250 may permit irradiation targets 250 to pass into a reservoir connector 1220 at desired times or at desired rates or amounts. For example, reservoir flow limiter 1250 may be a stop valve, gate valve, etc., that permits only a number of irradiation targets 250 that would be required to fill penetration tubing 1100 a length of L1 to enter reservoir connector 1220. Reservoir flow limiter 1250 may be operated directly or remotely or may be automatically programmed to dispense irradiation targets 250 at particular times and in particular fashions. Although a single irradiation target reservoir 1270, reservoir flow limiter 1250, and reservoir connector 1220 are shown connected to a loading junction 1200 and penetration tubing 1100 in FIG. 2, it is understood that more than one of these structures may be used. Further, these structures may be connected to multiple penetration pathways, such that a single reservoir 1270 may supply irradiation targets 250 into multiple penetration pathways and instrumentation tubes 50. Reservoir flow limiter 1250 may be further configured or programmed to direct irradiation targets 250 into appropriate reservoir connectors 1220 corresponding to individual penetration pathways and instrumentation tubes, if multiple reservoir connectors 1220 are used to ultimately reach multiple instrumentation tubes 50. Such an example may present an additional advantage of providing a single load point for irradiation targets 250 with access to multiple instrumentation tubes 50 for increased isotope generation. Irradiation target reservoir 1270 may connect to a loading junction 1200 via reservoir connector 1220 or any other pathway to provide irradiation targets 250 into loading junction 1200. Irradiation targets 250 may be moved between irradiation target reservoir 1270 and loading junction 1200 by any known mechanism, including reservoir flow limiter 1250, pneumatic force, magnetic force, gravity, etc. For example, if irradiation target reservoir 1270 connects to reservoir connector 1220 at a vertical height H2 above an entry point to loading junction 1200, gravity may drive irradiation targets 250 into loading junction 1200 if irradiation targets 250 are configured to move by gravity through reservoir connector 1220 or directly into loading junction 1200. If used, reservoir connector 1220 may be fabricated of a material and of a configuration that does not significantly react with or entrain materials from irradiation targets 250 coming into contact therewith, including, for example, aluminum, stainless steel, nickel alloys, PVC, PFA, carbon steel, rubber, etc. Reservoir connector 1220 may be cylindrical or any other shape that permits irradiation targets 250 to enter into and/or pass therethrough. For example, reservoir connector 1220 may have a generally circular cross section with a 0.5-inch diameter and smooth interior surface that permits spherical irradiation target 250 to roll between irradiation target reservoir 1270 and loading junction 1200. Loading and offloading systems useable in example embodiments permit irradiation targets to be loaded/harvested in a number of direction(s) based on their status and/or destination. Loading and offloading systems are operable during plant operation to properly load, guide, and harvest irradiation targets even when access to areas set off by access barrier 411 and instrumentation tubes 50 is limited. Any number of different sorting and/or directing mechanisms may be used as a loading and offloading system to achieve the desired movement of irradiation targets 250 within example embodiment systems. The example shown in FIG. 2 includes a loading junction 1200 capable of alternating between two paths: one path between penetration tubing 1100 and reservoir connector 1220; and another path between penetration tubing 1100 and retrieval path 1210. The example shown in FIG. 2 further includes a T-junction 1215 capable of alternating between two paths—one path between retrieval path 1210 and TIP tube 1310, and another path between retrieval path 1210 and harvesting cask 1290. Paths within loading junction 1200 may be configured and fabricated of a material that does not significantly react with or entrain materials from irradiation targets 250 coming into contact therewith. Loading junction 1200 may be embodied in several different ways. For example, apparatuses 400, 500, and/or 4100 disclosed in co-owned US Patent Publication 2011/0051875, Ser. No. 12/547,249, filed Aug. 25, 2009, incorporated by reference in its entirety, may be used for loading junction 1200 with appropriate configuration of pathways for accommodate irradiation targets 250 and connect with other elements of example embodiment irradiation target delivery and retrieval system 1000. Alternately, known devices for rerouting between pathways may be used for loading junction 1200, including diverters, turntables, sorters, etc. FIG. 3 is a detailed illustration of T-junction 1215 that may be used in loading and offloading systems. As shown in FIG. 3, retrieval path 1210, TIP tube 1310 and loading cask tube 1291 (or loading cask 1290 directly) may form a T-junction 1215 that permits one or more irradiation targets 250 to move therein, from retrieval path 1210 to cask tube 1291. FIG. 3 illustrates T-junction 1215 during an offloading or harvesting operation where plunger 1350 is withdrawn into TIP tube 1310 and deactivated so as to provide a path between retrieval path 1210 and harvesting cask 1290, such that irradiation targets 250 can move, by gravity, pneumatic force, mechanical action, magnetism, etc., into harvesting cask 1290. During a loading or other operation where access to harvesting cask 1290 is not desired, plunger 1350 may move beyond T-junction 1215 and into retrieval path 1210 so as to prevent any irradiation target 250 from moving into or out of harvesting cask 1290 or cask tube 1291, as shown by a position of plunger 1350 in FIG. 2. As shown in FIG. 3, a cask exhaust shaft 1281 and/or cask filter 1280 (FIG. 2) may be included on cask tube 1291, on retrieval path 1210, and/or elsewhere in example embodiments to provide an outlet for any pneumatic air used in example embodiments to escape and be filtered of airborne contaminants before exiting example embodiments. Cask exhaust shaft 1281 and/or cask filter 1280 may be configured through sizing or blockages to reduce or prevent irradiation targets 250 and isotopes produced in example systems from escaping from example systems. For example, if pneumatic driving system 500 is used in example embodiments to provide a pneumatic fluid flow and/or pressure to drive irradiation targets 250 and/or desired isotopes produced therefrom into cask 1290, cask exhaust shaft 1281 and/or cask filter 1280 may provide an exit for such air at atmospheric pressure while filtering the same. If used, retrieval path 1210 and/or cask tube 1291 may be fabricated of a material and of a configuration that does not significantly react with or entrain materials from irradiation targets 250 coming into contact therewith, including, for example, aluminum, stainless steel, nickel alloys, plastics, latex, etc. Retrieval path 1210 and/or cask tube 1291 may be cylindrical or any other shape that permits irradiation targets 250 to enter into and/or pass therethrough. For example, retrieval path 1210 and/or cask tube 1291 may have a generally circular cross section with a 0.5-inch diameter and smooth interior surface that permits spherical irradiation target 250 to roll therein. Oppositely, pathways not required to carry irradiation targets 250, such as cask exhaust shaft 1281 and/or TIPS tube 1310, may have smaller sizes, different cross-section shapes, and/or flow limiters not found in retrieval path 1210 and/or cask tube 1291 to prevent or reduce irradiation target 250 and isotopes produced therefrom from entering those structures. A plunger shield 1205 may be placed about retrieval path 1210 between 1200 and T-junction 1215. Plunger shield 1205 may be a radiation shielding structure that limits radiation exposure that may originate from plunger 1350 and cable 1305 in retrieval path 1210. For example, plunger shield may be any a solid material of a thickness and size placed around retrieval path 1210 to sufficiently block ionizing radiation. If plunger 1350 becomes radioactive, it may be stored in example embodiments inside of plunger shield 1205 to reduce worker exposure. As shown in FIG. 2, harvesting cask 1290 may be any storage receptacle capable of containing irradiated targets 250, including solid, liquid, and gaseous isotope products formed from irradiation targets 250 through irradiation. For example, harvesting cask 1290 may be a radiation-hardened shipping cask that complies with relevant radioactive shipping regulations. If harvesting cask 1290 is accessible outside of access barrier 411 at any time during operation of a nuclear power plant, it may be possible to relatively quickly harvest produced isotopes from example embodiments. Harvesting cask 1290, although shown as connected to a single example embodiment irradiation target delivery and retrieval system 1000 in FIG. 2, may be connected to multiple systems so as to provide a shared harvesting cask 1290 for multiple systems irradiating irradiation targets 250 across many instrumentation tubes 50. For example, multiple T-junctions 1215 may connect to an individual harvesting cask 1290 where all irradiation targets from multiple example embodiments may be stored and harvested. Of course, individual harvesting casks 1290 may be used for individual example embodiment irradiation target delivery and retrieval systems, to segregate irradiation targets 250 by individual system and/or instrumentation tube 50. In FIG. 2, harvesting cask 1290 and/or cask tube 1291 is shown equipped with a target counter 1295 that counts or measures properties of irradiation targets 250 moving into harvesting cask 1290. For example, target counter 1295 may count a number of discreet irradiated targets 250 that pass into harvesting cask 1290, using an optical or magnetic counter capable of counting each irradiated target 250. Similarly, target counter 1295 may detect and/or measure properties of irradiation targets 250 such as radioactivity, mass, temperature, flow rate, etc., that may be used to determine a number of irradiation targets 250, an amount of produced radioisotope, or any other property of irradiated targets 250. For example, target counter 1295 may be used to determine when all irradiated irradiation targets 250 and/or all expected produced isotope material has entered harvesting cask 1290 to indicate completion of a harvesting operation and readiness of harvesting cask 1290 to be removed to delivered to processing or end users. Although loading and offloading systems shown in the example of FIG. 2 include a loading junction 1200 and a T-junction 1215 to discriminate among multiple entry and offloading paths for irradiation targets 250 moving therethrough, it is understood that different apparatuses and paths may be used in example embodiments, depending on multiplicity of loading, offloading, and penetration points. For example, as shown in FIG. 4, loading junction 1200 may be capable of creating three or more paths into/out from penetration tubing 1100 to create discrete paths between penetration tubing and target reservoir 1270, TIP tube 1310, and harvesting cask 1290. Such a device includes device 500 from the incorporated 2011/0051875 document. Although FIG. 4 is shown in a loading configuration where irradiation targets 250 move from target reservoir 1270 through loading junction 1200 into penetration tubing 1100 for loading and irradiation into associated instrumentation tube 50, it is understood that loading junction is capable of separately or simultaneously providing a path for plunger 1350 to extend from retrieval path 1210 through loading junction 1200 into penetration tubing 1100 and a path from penetration tubing 1100 through loading junction 1200 into harvesting cask 1290. Still further, loading and offloading systems may include multiple T-junctions 1215 (shown singly in FIG. 2) arranged in series, or multiple guides (element 4100 from the incorporated 2011/0051875 document) arranged in series, to successively create additional paths that can be accessed or blocked as desired to create a desired origin and destination through example systems in nuclear power plants. Example embodiments may include drive systems or be operable with existing drive systems that move irradiation targets 250 from an origin, into an instrumentation tube 50, and to an accessible harvesting point following irradiation in example embodiment irradiation target delivery and retrieval systems. As shown in FIG. 2, at various points gravity may drive irradiation targets 250 between desired points in example systems. Similarly, pneumatic forces, such as those caused by induced pressure differentials within example systems, may move irradiation targets 250 within example systems. Systems providing magnetic or mechanical forces may similarly be used in example embodiments to move irradiation targets 250 therein. Additionally, a TIP drive 1300 may be used in example embodiments to move irradiation targets 250 in example system 1000. As shown in FIG. 2, a TIP drive 1300 may be used to supplement movement of irradiation targets 250 through example system 1000. TIP drive 1300 may be existing at some nuclear reactor plants or newly installed. For example, TIP drive 1300 may be drive system 300 from the incorporated 2011/0051875 document or another driving device. TIP drive 1300 may be communicatively connected to a control room or automatically or manually operable to drive irradiation targets 250 to known positions within example systems based on operation of TIP drive 1300. As shown in FIG. 2, TIP drive 1300 may include TIP tube 1310 that may connect TIP drive 1300 to loading/offloading systems in example embodiments where TIP drive 1300 may mechanically move irradiation targets through example systems. TIP Tube 1310 may be configured to pass through a TIP room wall 430 and any other structures to connect with example systems. As with other tubing useable in example embodiments, TIP drive 1300 may include a TIP filter 1320 that permits gas exhaust and/or pneumatic air flow out from example embodiment systems following a filtering to remove contaminants. As shown in FIG. 5, TIP drive 1300 may drive a helical TIP cable 1305. For example, TIP cable 1305 may resemble a rope having a relatively long driving portion. TIP cable 1305 may be made from a material having a low nuclear cross-section such as aluminum, silicon, stainless steel, etc. and may be braided in order to increase flexibility, stiffness and/or strength so that TIP cable 1305 may be easily bendable and capable of snaking through penetration pathways and loading/offloading systems of example embodiments. Although TIP cable 1305 may be flexible, it also has sufficient axial rigidity to transmit a driving mechanical force without buckling. TIP cable 1305 may further have a diameter and other physical characteristics configured to fit and move within any tubing or other transit portions of example embodiments and instrumentation tube 50. TIP cable 1305 may include a helical winding configured to cooperate with a helical gear that may be present in TIP drive 1300. As shown in FIG. 5, plunger 1350 may be connected to TIP cable 1305 at a connection point 1315. For example, an existing TIP cable 1305 may be modified by removing instrumentation or another feature on its end and attaching plunger 1350. Plunger 1350 may be relatively rigid and include a head portion capable of bearing or pushing irradiation targets 250 through example systems. Plunger 1350 may be fabricated of a material that substantially maintains its physical characteristics within an operating nuclear reactor environment while not substantially reacting with or entraining any irradiation target 250 or isotope produced therefrom. For example, plunger 1350 may be fabricated of a steel or aluminum alloy, high-temperature ceramic, etc. Plunger 1350 may further be formed of a ferromagnetic material or sensor such that its position may be tracked with appropriate detectors in example embodiments, such as sensors in plunger shield 1205, flow limiter 1105, etc. Plunger 1350 may further be shaped and sized to fit nearly flush with TIP tube 1310 so as to prevent or reduce any irradiation target 250 or isotope product movement into TIP tube 1310 past loading/offloading systems. Example drive systems may be pre-existing and configured and/or newly installed and useable with example embodiment irradiation target delivery and retrieval system 1000. For example, an existing TIP system may be removed past TIP tube 1310 and connected to example system as shown in FIG. 2 installed in a nuclear facility, or a driver system may be newly installed and connected to example embodiment irradiation target delivery and retrieval system. If example systems are connected by tubing or other enclosed transit paths, a pneumatic driving system 500 and/or other pneumatic force devices may be used to drive irradiation targets 250 therein while filtering all pneumatic exhaust through HEPA-type filters at exhaust points so as to reduce or prevent escape of produced isotope contaminants. It is further understood that a drive system useable with example embodiments may be entirely passive or gravity-driven, by appropriately choosing height H2 and relative positioning of irradiation target 250 origins and harvesting terminals. Irradiation targets 250 are capable of movement, including rolling, conveyance, flow, etc., within example embodiment irradiation target delivery and retrieval systems and, as such, may take on many physical forms. Irradiation targets 250 further substantially convert to a desired isotope after being exposed to neutron and/or other radiation within reactor vessel 10 and/or core 15. FIGS. 6A and 6B are illustrations of example embodiment irradiation targets 250a and 250b in generally spherical form. As shown in FIG. 6A, irradiation target 250a may be a solid sphere of material that readily converts to desired isotopes after hours or days of exposure to radiation generated within nuclear core 15. Alternatively, as shown in FIG. 6B, irradiation target 250b may include a containment shell 251 that houses a solid, liquid, or gaseous material 252 that converts to desired isotopes which are equally contained by containment shell 251 after hours or days of exposure to radiation from nuclear core 15. One potential advantage of using example embodiment irradiation targets 250a and 250b is that the spherical surface may be polished and made even so as to readily roll within tubing useable in example embodiments, reducing friction and risk of snagging, damage, or abrading to irradiation targets 250. Of course, other shapes and configurations of irradiation targets 250 are useable in example embodiments to match other types of penetration pathways, instrumentation tube configurations, and loading/offloading systems. The materials used in irradiation targets 250 may be selected from any material having an absorption cross-section for radiation encountered in instrumentation tube 50 that will permit the material to appreciably convert to a desired daughter product. For example, cobalt-59, iridium-191, iridium-193, tungsten-185, tungsten-187, etc. may be converted to valuable precious metals. Similarly, molybdenum-98, dysprosium-164, holmium-165, lutetium-176, potassium-41, sodium-23, and ytterbium-168 may be converted into valuable radioisotopes useable in various known medical imaging, scanning, diagnostic, and therapeutic treatments. One particular advantage possible with example embodiments and methods is that these latter radioisotopes, each having shorter half-lives on the scale of hours or days, may be readily generated and harvestable without delays due to inaccessibility of access restricted areas proximate to instrumentation tube openings in nuclear power plants. As a specific example, irradiation targets 250 may be fabricated as spheres of molybdenum-98 or naturally occurring molybdenum, converted to molybdenum-99 through a few hundred hours of irradiation with thermal neutron flux in an operating nuclear reactor, and immediately deposited in a harvesting cask where the molybdenum-99, with a half-life of 66 hours, will decay with a half-life of 66 hours into technetium-99 with a half-life of 6 hours. The harvesting cask may be transported to medical facilities potentially around the world while the molybdenum-99 is decaying, presenting the medical facilities with a ready source of technetium-99. Irradiation targets 250 may further include a tracking target located at a known position among all other targets that is fabricated of a material that is different from all other targets and permits tracking or locating of irradiation targets 250. For example, the first and last irradiation target 250 may be fabricated of a ferromagnetic material that can be tracked with a magnet sensor, or may be fabricated of a material converting to a different isotope product that can be detected with a radiation sensor. Or, for example, irradiation targets 250 may be fabricated of a material and mass corresponding to an amount of radiation expected to be encountered at a known axial level of the irradiation target 250. By varying the detectable physical properties of irradiation targets 250, it may be possible to ensure that all targets have moved properly through example systems and/or that an optimal amount of desired isotope produce is produced from irradiation targets 205. Example Methods Example embodiment irradiation target delivery and retrieval systems are useable with example methods to load, irradiate, and harvest irradiation targets in a continuous and timely fashion within an otherwise potentially inaccessible nuclear reactor. FIG. 8 is a flow chart of an example method that may be carried out with example embodiments. As shown in FIG. 8, one or more irradiation targets may be supplied to an irradiation target delivery and retrieval system within a nuclear power plant in S100. Supplying may include a periodic or continuous supplying to a cache of targets or supplying only an amount for immediate use in example methods. Different types and numbers of irradiation targets may be supplied in S100, and supplying may be achieved during commercial operation of the nuclear power plant. For example, using the example system of FIG. 2, a number of irradiation targets 250 required to fill a number of instrumentation tubes 50 may be supplied to a shared irradiation target reservoir 1270 or to multiple target reservoirs 1270 for each instrumentation tube and associated system. In S200, a penetration pathway is created by loading/offloading systems to permit irradiation targets to move from the supply point of S100 into an access-restricted area and one or more instrumentation tubes for irradiation. For example, as shown in FIG. 2, loading junction 1200 may create a pathway for irradiation targets 250 to pass from reservoir 1270 into penetration tubing 1100 by rotating or otherwise sealing off other paths. Reservoir flow limiter 1250 releases a desired number of irradiation targets 250 into a desired loading junction 1200 (among several potential junction points if a shared reservoir is used) and penetration tubing 1100. As discussed above, several different path creation mechanisms may be utilized in S200, including advancing plunger 1350 to a T-junction (not shown) between retrieval path 1210, penetration tubing 1100, and irradiation target reservoir 1270 to create a single path for irradiation targets 250 into penetration tubing 1100. Irradiation targets 250 may move by gravitational, pneumatic, magnetic, mechanical, etc. forces into penetration pathways in S200. In S250, irradiation targets may be held outside an access barrier while loading/offloading systems seal off reservoirs, create different flow paths, and/or permit a driving system to access irradiation targets. For example, as shown in FIG. 2, irradiation targets 250 may be stopped by pathway flow limiter 1105 and back up to loading junction 1200 a distance of L1, and loading junction 1200 may create a different path such that plunger 1350 and TIP cable 1305 can extend out of TIP tube 1310, through retrieval path 1210 and loading junction 1200 to bias against a final irradiation target 250 in penetration tubing 1100. It is understood that, like any step, S250, may be omitted, particularly in the instance of a purely gravitational-driven system. In S300, irradiation targets move through penetration pathways into an access-restricted area such as drywell 20 and a corresponding instrumentation tube. As shown in FIG. 2, irradiation targets 250 may move through penetration tubing 1100 through access barrier 411 and around or through any structures therein, such as pedestal 412, to flange 1110 and into instrumentation tube 50 spatially inside reactor vessel 10. In S400, irradiation targets are held within an instrumentation tube at desired positions and for a desired amount of time to generate daughter isotopes from exposure to radiation within instrumentation tubes. As shown in FIG. 7, irradiation targets 250 may extend axially in instrumentation tube 50 a length of core 15 (approximately 12 feet for a some conventional BWR cores), absorbing neutron flux from core 15 and converting into desired isotopes. In the example of FIG. 7, irradiation targets 250 are maintained by plunger 1350 extending all through penetration tubing 1100 and maintained at flange 1110 to hold irradiation targets 250 at axial positions within instrumentation tube 50 until irradiation is complete. It is understood that other holding mechanisms, including valves, latches, gravity form counterbalancing, etc. may equally maintain irradiation targets in S400. In S500, irradiation targets are removed from the instrumentation tube after residence in the reactor for a predetermined time for producing a desired amount of harvestable isotope. For example, plunger 1350 and TIP cable 1305 may be withdrawn out through access barrier 411 and irradiated irradiation targets 250 may descend by gravity or any other force out of instrumentation tube 50. In S550 additional driving forces may be used to move irradiation targets back out of penetration pathways. For example, pneumatic driving system 500 may force irradiation targets 250 out of an area restricted by access barrier 411 through a pressure differential formed from injected air. Of course, other mechanical, magnetic, electrical, etc. forces may be used to drive irradiation targets 250 through penetration tubing 1100 outside of access barrier 411. In S600, at some point following insertion of irradiation targets in S300, loading/offloading systems are reconfigured to provide an exit pathway for irradiation targets to be harvested following irradiation. For example, as shown in FIG. 4, plunger 1350 may be withdrawn through loading junction 1200 to T-junction 1215 to form a pathway into harvesting cask 1290. Alternatively, in an example system as shown in FIG. 4, loading junction 1200 may rotate to or otherwise create a unique path into harvesting cask 1290 for irradiated targets 250, following withdrawal of plunger 1350. Of course, multiple T-junctions and other combinations of forks, dividers, turntables, sorters, etc., may be used to create the unique loading and harvesting paths in loading/offloading systems usable with example methods and embodiments. In S700, the irradiated irradiation targets containing desired isotopes produced therein are moved through delivery and retrieval systems to an accessible exit for shipping or consumption, regardless of plant state. For example, as shown in FIG. 3, irradiation targets 250 may be directed into harvesting cask 1290 by gravity and/or pneumatic forces. Harvesting cask 1290 may then be used as shipping cask or emptied into a replacement shipping cask and shipped directly to end users. A single or multiple exits may be used in S700, with irradiation targets from several example systems and instrumentation tubes being emptied into a single harvesting point or individual, separate points. Example methods may be executed continuously and automatically without significant user interaction, based on product need, plant conditions, and/or irradiation target supply. For example, various sensors, flow limiters, and drives of example system 1000 may be used to coordinate movement of irradiation targets 250 through example system 1000. Irradiation targets 250 may be supplied in type and amount to particular systems and instrumentation tubes 50 from single or multiple target reservoirs 1270, and their harvesting into one or more harvesting casks 1290 may be equally coordinated to ensure a desired number and type, including activity and daughter product identity, are present in specific harvesting casks 1290. Further, users or plant operators may be able to remotely track or control operations of example systems through communicative connections with the various sensors, flow limiters, and drives of example system 1000. It is understood that creation of pathways from various points in example systems and example methods may include the installation of tubing or other pathways and pathway selection systems including loading/offloading systems within access-restricted areas and nuclear plant facilities when accessible, such as during a plant outage. For example, penetration tubing 1100, reservoir tubing 1220, and/or retrieval path 1210 may all be installed as parts of example methods. Example embodiments and methods thus being described, it will be appreciated by one skilled in the art that example embodiments may be varied and substituted through routine experimentation while still falling within the scope of the following claims. For example, the types and numbers of penetration pathways, loading/offloading systems, and drive systems falling within the claims are not limited to the specific systems shown and described in the figures—other specific devices and systems for loading irradiation targets into an access-restricted area of a nuclear power station and instrumentation tube for irradiation and offloading the same outside the access-restricted area for harvesting are equally useable as example embodiments and fall within the scope of the claims. Such variations are not to be regarded as departure from the scope of the following claims.
051568187	claims	1. Apparatus for isolating hazardous radioactive waste for disposal comprising: a) a bifurcated centrifugal casting mold having at least two separable mold parts, said mold being supported for rotation about a first axis; b) means for supporting a completed monolith in said apparatus with the mold parts removed therefrom; c) powered drive means for rotating said mold and said monolith about said first axis; d) mold removal means aligned along a second axis substantially perpendicular to said first axis for removing the separate parts of said bifurcated casting mold from a monolith while leaving said monolith supported in said apparatus for rotation about said first axis; e) means for injecting a charge of radiation shielding material into a pre-formed shell placed in said mold; and f) means for heating the interior of said shell during rotation of said mold about said first axis. 2. The apparatus of claim 1, wherein said means for supporting comprises a gripping chuck and a retaining collar each having an aperture for receiving an axially extending hub on said shell, said gripping chuck and said retaining collar being moveable toward and away from each other along said first axis into and out of engagement with said hubs for supporting said shell. 3. The apparatus of claim 1, wherein said mold removal means comprises: a pair of linear actuators, each of said actuators having a mold engaging end which is removeably attachable to a respective one of said mold halves for supporting said mold halves when said actuators are attached thereto. 4. The apparatus of claim 3, further comprising a threaded connector affixed to the mold engaging end of each of said actuators, said mold halves each have a threaded connector affixed thereto for engagement with said actuator connectors, and means for rotating said actuator connectors. 5. The apparatus of claim 1, wherein said means for injecting comprises: an elongated generally cylindrical feed wand receivable in said shell through a fluid inlet port, said wand having a distal end which is receivable in a fluid outlet port to block said outlet port. 6. The apparatus of claim 5, further comprising means for moving said feed wand longitudinally into and out of said shell. 7. The apparatus of claim 5, wherein said means for heating comprises an elongated electrical radiation heater disposed in said shell and extending axially therein from end to end of said shell for blocking both said fluid inlet port and said fluid outlet port during biaxial rotation of said shell during curing of said radiation shielding material therein.
039363490	summary	BACKGROUND OF THE INVENTION This invention relates to nuclear reactors of the kind having a core composed of closely packed parallel fuel elements which are provided with bearing pads at selected levels whereby the elements are firmly positioned against one another to prevent them rattling. Such a nuclear reactor core is disclosed in U.S. Pat. No. 3,383,287 and the fuel elements in that reactor core are arranged in groups in each of which each element is supported in such a manner as to tend to tilt the element towards the centre of the respective group whereby the elements of each group are urged laterally into firm engagement either with one another or with a structural member around which the group may be arranged. Thus the elements of each group are jammed together, and, where desirable, about a central post on which the surrounding elements lean. SUMMARY OF THE INVENTION According to the present invention, in a nuclear reactor core comprising an array of closely packed components which include fuel elements, the components being arranged with their longitudinal axes generally vertical in groups, at least some of the components in each of several groups are supported in such manner as to tend to tilt the components towards the centre of the respective group whereby the components of each group are urged laterally into abutment with one another, and the fuel elements have interlocking bearing pads intermediate their ends whereby relative lateral movement of two adjacent fuel elements is resisted. By reducing sideways slip of fuel elements the stability of of the reactor core is improved during refuelling operations. The invention also resides in a nuclear reactor fuel element for a nuclear reactor core in accordance with the preceding paragraph, the fuel element comprising a cluster of parallel fuel pins enclosed by a peripheral wrapper of hexagonal cross-section, the bearing pads being in the form of spline-like ribs extending parallel to the longitudinal axis of the fuel element and interlockable with complementary ribs of pads on adjacent fuel elements in the group. The invention is applicable to a nuclear reactor core wherein the components, which comprise fuel elements and control rods, are generally arranged in modules each module comprising a cluster of four components at least three of which are fuel elements, one fuel element being rigidly supported whilst the remaining components are resiliently tilted towards the centre of the cluster to lean on the rigidly supported element. The invention is also applicable to a nuclear reactor core wherein the fuel elements are arranged in modules, each module comprising a cluster of six fuel elements, each resiliently tilted towards a central void to form a circular arch. The modules may include additional fuel elements disposed outside the cluster and resiliently tilted towards the central void. The central void may be used to accommodate a control rod. By using a module arrangement of fuel elements wherein one element serves to support the remaining elements and control rods of the cluster, or wherein the elements lean on each other in a circular arch, the need for structural members to form leaning posts in the manner described in U.S. Pat. No. 3,383,287 is avoided. Such structural members are subject to irradiation embrittlement which could lead to failure within the core.
063109344	abstract	An X-ray projection exposure apparatus includes a mask chuck for holding a reflection X-ray mask having a mask pattern thereon, a void being formed between the mask and the mask chuck, a wafer chuck for holding a wafer onto which the mask pattern is transferred, an X-ray illuminating system for illuminating the reflection X-ray mask, held by the mask chuck, with X-rays, an X-ray projection optical system for projecting the mask pattern of the reflection X-ray mask onto the wafer held by the wafer chuck with a predetermined magnification and a supply for supplying the void formed between the mask and the mask chuck with a cooling gas for cooling the mask.
	claims	1. An apparatus configured to inspect at least a portion of a top guide of a nuclear reactor, wherein the nuclear reactor comprises a plurality of beams arranged in a grid pattern, the apparatus comprising:an alignment assembly comprising a plurality of legs and an actuator configured to move the plurality of legs between a retracted position wherein at least one of the legs of the plurality of legs is disengaged from at least one beam of the plurality of beams and an extended position wherein all of the legs of the plurality of legs are engaged with at least one beam of the plurality of beams; andan inspection element coupled to the alignment assembly, wherein the inspection element is configured to be positioned a predetermined distance away from a beam of the plurality of beams when the legs are in the extended position. 2. The apparatus of claim 1, further comprising a pair of supports, wherein each support of the pair of supports comprises an engagement edge configured to mechanically engage an upper edge of a beam of the plurality of beams. 3. The apparatus of claim 1, further comprising a plurality of links that extend between the actuator and at least one leg of the plurality of legs. 4. The apparatus of claim 1, further comprising a drive apparatus, wherein the drive apparatus is coupled to the inspection element, and wherein the drive apparatus is configured to translate the inspection element between a first location and a second location. 5. The apparatus of claim 4, wherein the first location is proximal a first beam of the plurality of beams, and wherein the second location is proximal a second beam of the plurality of beams. 6. The apparatus of claim 4, further comprising a second inspection element. 7. The apparatus of claim 6, wherein the drive apparatus is coupled to the second inspection element, and wherein the drive apparatus is configured to translate the second inspection element between a first location and a second location. 8. The apparatus of claim 7, wherein the third location is proximal a first beam of the plurality of beams, and wherein the fourth location is proximal a second beam of the plurality of beams. 9. The apparatus of claim 1, wherein each leg of the plurality of legs comprises an L-shaped configuration comprising a first portion extending in a first direction and a second portion extending in a second direction, wherein the first direction is different than the second direction. 10. The apparatus of claim 9, wherein the first direction extends generally away from the actuator, and wherein the second direction extends generally towards the plurality of beams. 11. The apparatus of claim 1, further comprising an extension mechanism configured to translate the inspection element between a first position relative to the plurality of beams and a second position relative to the plurality of beams. 12. The apparatus of claim 11, wherein the inspection element is positioned a first distance away from the plurality of beams in the first position, wherein the inspection element is positioned a second distance away from the plurality of beams in the second position, and wherein the first distance is different than the second distance. 13. The apparatus of claim 1, wherein the inspection element is configured to emit an ultrasonic wave towards the top guide of the nuclear reactor. 14. An alignment assembly configured to properly align an inspection element relative to a top guide of a nuclear reactor, wherein the top guide comprises a plurality of beams arranged in a grid pattern, wherein each beam of the plurality of beams comprises an upper edge, the alignment assembly comprising:a plurality of legs; andan actuator configured to move the plurality of legs between a retracted position and an extended position, wherein at least one of the legs of the plurality of legs is disengaged from a beam of the plurality of beams in the retracted position, wherein at least one of the legs of the plurality of legs engages a beam of the plurality of beams in the extended position. 15. The alignment assembly of claim 14, further comprising a pair of supports, wherein each support of the pair of supports comprises an engagement edge configured to mechanically engage an upper edge of a beam of the plurality of beams. 16. The alignment assembly of claim 14, further comprising a plurality of links that extend between the actuator and at least one leg of the plurality of legs. 17. The alignment assembly of claim 14, further comprising a drive apparatus, wherein the drive apparatus is configured to be coupled to the inspection element, and wherein the drive apparatus is configured to translate the inspection element between a first location and a second location. 18. The alignment assembly of claim 17, wherein the first location is proximal a first beam of the plurality of beams, and wherein the second location is proximal a second beam of the plurality of beams. 19. The alignment assembly of claim 14, further comprising an extension mechanism configured to translate the inspection element between a first position relative to the plurality of beams and a second position relative to the plurality of beams. 20. The alignment assembly of claim 19, wherein the inspection element is positioned a first distance away from the plurality of beams in the first position, wherein the inspection element is positioned a second distance away from the plurality of beams in the second position, and wherein the first distance is different than the second distance. 21. The apparatus of claim 14, wherein the inspection element is configured to emit an ultrasonic wave towards the top guide of the nuclear reactor.
	claims	1. A device for applying laser radiation (13) to an outside surface of a rotationally symmetrical component (11, 11′, 11″), comprising a plurality of lenses (10), through which the laser radiation (13) passes and which are designed and/or are arranged in such a way that an axis of symmetry (12, 12′, 12″) of the component (11, 11′, 11″) is located at a focal point of each of the lenses (10). 2. The device according to claim 1, wherein the device comprises homogenizers (14, 18, 21), one homogenizer associated with each of the lenses (10), homogenizing the laser radiation (13) before passing through each of the lenses (10). 3. The device according to claim 2, wherein when a number of the lenses (10) is less than eight, the homogenizers (14, 18, 21) are designed such that the laser radiation (13) has a top-hat intensity distribution wherein intensity decreases toward edges according to cos 2 (φ), and wherein, when a number of the lenses (10) is greater than or equal to eight, the homogenizers (14, 18, 21) are designed such that the laser radiation (13) has a linear top-hat intensity distribution. 4. The device according to claim 2, wherein the homogenizers (14,18) comprise at least one lens array (15, 16) or a light guide (19). 5. The device according to claim 4, wherein the homogenizers (14, 18) comprise two lens arrays (15, 16). 6. The device according to claim 2, wherein the homogenizer (21) comprises a waveguide (23) having at least partially cuboidal light-guiding region (26). 7. The device according to claim 6, wherein a first extent (D1) of the light-guiding region (26) that extends between an entrance face and an exit face of the light-guiding region (26) is greater than a second (D2) and/or a third extent (D3) that is perpendicular to the first extent (D1) of the light-guiding region (26). 8. The device according to claim 7, wherein first extent (D1) of the light-guiding region (26) that extends between an entrance face and an exit face of the light-guiding region (26) is greater than a second and/or third extent (D2, D3) that is perpendicular to the first extent (D1) of the light-guiding region (260), by a factor of at least 3. 9. The device according to claim 8, wherein the factor is between at least 7 and at most 100. 10. The device according to claim 2, wherein the device comprises an optical arrangement to expose at least one of the lenses (10) to laser radiation (13). 11. The device according to claim 10, wherein the optical arrangement and/or the homogenizer (14, 18, 21) are designed such that an entrance face of each of the lenses (10) is exposed to a homogeneous spatial distribution or to a homogeneous spatial intensity distribution of the laser radiation (13). 12. The device according to claim 10, wherein the optical arrangement and/or the homogenizer (14, 18, 21) are designed such that an entrance face of each of the lenses (10) is exposed to a homogeneous spatial distribution or to a homogeneous spatial intensity distribution of the laser radiation (13) matching a design of the lens (10). 13. The device according to claim 10, wherein the homogenizers (14, 18, 21) are encompassed by the optical arrangement. 14. The device according to claim 10, wherein the optical arrangement comprises a collimator. 15. The device according to claim 14, wherein the collimator comprises at least one lens for collimation, which is designed and arranged in the device such that the laser radiation is incident on the lenses (10) with no divergence or with the smallest possible divergence, or is at least substantially collimated. 16. The device according to claim 2, wherein the lenses (10) are arranged in the circumferential direction next to one another. 17. The device according to claim 16, wherein the lenses (10) adjoin one another, with respect to the axis of symmetry (12, 12′, 12″) of the component (11, 11′, 11″). 18. The device according to claim 2, wherein at least one of the lenses (10) are arranged so that the direction in which the laser radiation (13) propagates after passing through the at least one of the lenses (10) has with respect to the axis of symmetry (12, 12′, 12″) of the component (11, 11′, 11″) both a radial and an axial component. 19. The device according to claim 2, wherein each of the lenses (10) is arranged lenses (10) has with respect to the axis of symmetry (12, 12′, 12″) of the component (11, 11′, 11″) both a radial and an axial component. 20. The device according to claim 1, wherein each of the lenses (10) has for the laser radiation (13) an enhance face facing away from the component (11, 11′, 11″) and an exit face facing the component (11, 11′, 11″). 21. The device according to claim 1, wherein at least two the lenses (10) have an identical focal length (f). 22. The device according to claim 1, wherein at least one of the lenses (10) are cylindrical lenses having cylinder axes that are parallel to the axis of symmetry (12, 12′, 12″) of the component (11, 11′, 11″). 23. The device according to claim 1, wherein at least one of the lenses (10) have a rotationally-symmetrical curvature. 24. The device according to claim 23, wherein at least one of the lenses (10) have a rotationally-symmetrical curvature. 25. The device according to claim 24, wherein at least one of the lenses (10) are spherical lenses (10). 26. The device according to claim 24, wherein the lenses (10) are spherical lenses (10). 27. The device according to claim 1, wherein each of the lenses (10) has an identical focal length (f). 28. The device according to claim 1, wherein the device comprises an optical arrangement to expose each of the lenses (10) to laser radiation (13). 29. The device according to claim 1, wherein the lenses (10) are cylindrical lenses having cylinder axes that are parallel to the axis of symmetry (12, 12′, 12″) of the component (11, 11′, 11″). 30. The device according to claim 1, wherein each of the lenses (10) has a rotationally-symmetrical curvature. 31. The device according to claim 1, wherein the plural lenses (10) are three or more lenses arranged in a closed geometric configuration with the plural lenses (10) adjoining at corners of the closed geometric configuration with the component (11, 11′, 11″) enclosed by the geometric configuration. 32. The device according to claim 1, wherein none of the laser radiation bypasses the component (11, 11′, 11″). 33. The device according to claim 1, wherein an entire angular space of a circumference of the component (11, 11′, 11″) is illuminated, and wherein a numerical aperture of the plural lenses (10) is determined as follows:NA=sin(α/2)with α=360°/number of lenses.
048715085	summary	BACKGROUND OF THE INVENTION A core in a boiling reactor normally comprises several hundred fuel assemblies. These are arranged vertically in the core and have an at least substantially square cross-section. Each fuel assembly consists of a bundle of fuel rods which are surrounded by a fuel channel. The core also comprises a large number of control rods, which may each comprise four vertical blades arranged in a cruciform configuration and provided with a neutron absorber material. The fuel assemblies are normally arranged in a symmetrical lattice, with each fuel assembly included in two rows of fuel assemblies, located perpendicular to each other, and the control rods normally being arranged with each one of their absorber blades between two fuel assemblies located in the same row, so that each control rod together with four fuel assemblies, arranged around its blades, forms a unit, in the following called control rod unit, having an at least substantially square cross-section and so that the control rod units are arranged in a symmetrical lattice, with each control rod unit included in two rows of control rod units located perpendicular to each other. The present invention relates to the operation of a nuclear reactor of the type described above. When the burnup in a reactor has progressed so far that the smallest acceptable core reactivity margin has been attained, a partial recharging of fuel is carried out. By suitably balancing the amount of fuel to be replaced as well as the fissile enrichment of the replacement fuel, an excess reactivity is effected which permits a certain energy output until the next refuelling occasion. In connection with refuelling, when the reactor is shut down, an exchange of control rods is also carried out, if necessary, in which control rods the reactivity worth (by consumption of neutron absorber) has dropped to a predetermined value, or which control rods exhibit defects. Normally, the exchange of control rods is carried out if the reactivity worth in one-fourth of the length of the control rod has dropped to 90% of the original reactivity worth. By reactivity worth of a control rod is meant that change in reactivity which follows the complete insertion of a fully withdrawn control rod in a critical reactor under given conditions. When exchanging control rods, control rods existing in the reactor at the time of the exchange are replaced, according to the state of the art, by control rods having the same reactivity worth as that of the existing ones in their original state. In this way, the behaviour of the reactor is maintained unchanged. The control rods in a boiling reactor bind a certain amount of reactivity in the reactor. Together with burnable neutron absorbers, for example digadolinium trioxide, in the fuel, a smallest permissible shutdown margin is obtained for a given core. If it is desired to achieve longer fuel cycles, i.e., a longer operating time between two refuellings, or a larger energy output in the reactor, the shutdown margin in a core with existing control rods has to be dimensioned using an increased quantity of burnable neutron absorber. Such use of an increased quantity of burnable absorber results in reactivity losses because of increased reactivity penalty of residual burnable absorber at the end of a fuel cycle (such as because of an increased content of isotopes of gadolinium with a low absorption cross-section for neutrons). Thus, an increased use of a burnable absorber enables a greater flexibility regarding the design of the core with reactor fuel but at the same time results in increased losses because of the reactivity penalty of residual neutron absorber. SUMMARY OF THE INVENTION The present invention is based on the realization that the increased shutdown margin, which is required for longer fuel cycles or for a greater energy output, can be brought about without the above-described disadvantages--consisting of the increased reactivity penalty caused by the use of an increased amount of neutron absorbers--occurring. According to the invention, this is achieved by the fact that some of the control rods located in the reactor, when exchanging control rods, are replaced by control rods having a higher reactivity worth than the original reactivity worth of the control rods present in the reactor at the time of exchange, whereas other control rods used during the earlier operation of the reactor are maintained in the reactor. The invention makes it possible, with a limited number of control rods with a higher reactivity worth, to achieve the same operational advantages as if all the existing control rods were to be exchanged for others with a higher reactivity worth. The advantages thus gained are very considerable. The present invention relates more particularly to a method for the operation of a light water boiling reactor with a core comprising a plurality of vertical fuel assemblies having an at least substantially square cross-section, each fuel assembly consisting of a bundle of fuel rods surrounded by a fuel channel, and a plurality of control rods, each control rod comprising four vertical blades arranged in a cruciform and provided with a neutron absorber material, the fuel assemblies being arranged in a symmetrical lattice with each fuel assembly included in two rows of fuel assemblies which are perpendicular to each other and the control rods being arranged with each one of their blades between two fuel assemblies located in the same row, so that each control rod together with four fuel assemblies, arranged around the blades of the control rod, forms one unit, a control rod unit, having an at least substantially square cross-section, the control rod units being arranged in a symmetrical lattice with each control rod unit included in two rows of control rod units perpendicular to each other, characterized in that, after a period of operation of the reactor, when exchanging control rods, which are present in the reactor at the time of exchange and which have been used during the operating period, for new control rods, there are arranged in some control rod units control rods having a reactivity worth which is higher than the original reactivity worth of the control rods which are present in the reactor at the time of exchange and have been used during the operating period, whereas in other control rod units there are used control rods which have been used in the reactor during the operating period. The reactivity worth, in a cold shutdown reactor, of the newly inserted control rods with the higher reactivity worth is at least 6% higher, preferably 10-20% higher, than the original reactivity worth of the control rods present in the reactor at the time of the exchange. By original reactivity worth of a control rod is meant in this patent application the reactivity worth of a control rod in a non-used state. To utilize the invention in full, the exchange of existing control rods for new control rods with a higher reactivity worth only takes place in control rod units located in a central zone of the reactor core, which is located inside an edge zone extending around the reactor core and comprising those control rod units which are located furthest out in the reactor core in each row of control rod units. According to one embodiment of the invention, control rods which have been used during a preceding operating period are maintained in a number of control rod units distributed over the central zone of the reactor core, whereas at least the control rods in those control rod units which are located adjacent to each such control rod unit with maintained control rod and in the same rows perpendicular to each other as this control rod unit with maintained control rod are replaced by new control rods with a higher reactivity worth. Control rods used during the earlier operating period in regions within the central part of the reactor core, comprising 3.times.3 control rod units, can then be exchanged for new control rods with a higher reactivity worth in that control rod unit which is located in the centre of the region and in those four control rod units which are located in the same rows, perpendicular to each other, as the control rod unit located in the centre, whereas control rods used during the earlier operating period are maintained in the remaining four control rod units. The exchange of fuel rods is preferably performed such that new control rods with a higher reactivity worth are arranged in three control rod units at most, located adjacent to each other, in the same row of control rod units. The number of control rods used during the earlier operating period, which are replaced by new control rods, preferably amounts to 40-60% of the number of control rods in the entire reactor core and to 50-80% of the number of control rods in the central zone of the reactor core.
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054886449	description	BEST MODE FOR CARRYING OUT THE INVENTION Referring now to FIG. 1, there is illustrated a nuclear fuel assembly, generally designated 20, including a plurality of fuel elements or rods 21 supported between an upper tie plate 22 and a lower tie plate 23. Fuel rods 21 pass through a plurality of fuel rod spacers 24a and 24b which provide intermediate support to retain the elongated rods 21 in spaced relation and to restrain them from lateral vibration. In one embodiment, seven spacers are elevationally evenly spaced along the fuel assembly. While an 8.times.8 array of fuel rods is disclosed, it will be appreciated that the invention hereof is applicable to arrays of different numbers, for example, matrices of 10.times.10 arrays. Each fuel rod 21 is formed of an elongated tubular cladding material with the nuclear fuel and other materials sealed in the tube by upper and lower end plugs 26 and 27, respectively. The lower end plugs 27 are registered and positioned in bores formed in the lower tie plate 23 for supporting the fuel bundle. The upper end plugs 26 are disposed in cavities 31 in the upper tie plate 22. Typically, the fuel rod assembly includes a channel 33 of substantially square cross-section sized to form a sliding fit over the upper and lower tie plates 22 and 23 and the spacers 24 so that the channel 33 may readily be remounted and removed. The lower tie plate 23 includes a nose piece 38 adapted to support the fuel assembly 20 in a socket in a core support plate (not shown) in the reactor pressure vessel. The end of the nose piece is formed with openings 39 to receive coolant/moderator for flow upwardly along and among the fuel rods. Turning now to FIG. 2, there is illustrated a spacer constructed in accordance with the present invention having the spring assemblies and ferrule combinations hereof. The spacer is comprised of a plurality of cylindrical ferrules 50 and spring assemblies 52, both of which are described in detail hereinafter. Each spacer 24 thus comprises a plurality of ferrules 50 arranged in a square matrix in which each ferrule receives a fuel rod and maintains the fuel rods spaced and restrained relative to one another. The spring assemblies 52 bias the fuel rods in a lateral direction against stops 54 whereby the fuel rods are maintained in a predetermined position relative to one another and in the spacer 24. Referring now to FIGS. 3-7, the spring assembly 52 will be described. Referring to FIG. 3, each spring assembly 52 includes two identical spring bodies, one of which is illustrated in FIG. 3. Thus, in FIG. 3, spring body 56a includes, in an unstressed condition, a central elongated leaf 58, a pair of outer leaves 60 straddling central leaf 58 and spaced laterally therefrom, and end portions 62 joining the ends of the central and outer leaves 58 and 60, respectively, at opposite ends of the spring body 56a. Central leaf 58 includes an arcuate projection 64 intermediate its ends and preferably medially thereof. Central leaf 58 also includes leg portions 66 which join the intermediate projection 64 and the end portions 62. Outer leaves 60 include intermediate, preferably medial, portions 68 which project in a direction opposite from the central portion 64 of the central leaf 58. Outer leaves 60 also include leg portions 70 joining the central portions 68 and the end portions 62 to one another. As best seen in FIG. 5, the leg portions 66 of central leaf 58 and leg portions 70 of outer leaves 60, as well as the end portions 62 of the spring body 56a lie in a common plane, designated A--A in FIG. 5 fin an unstressed condition of the spring assembly. Thus, the arcuate projection 64 of the central leaf 58 projects to a forward side of the plane A--A. More particularly, central portion 68 projects outwardly of the plane A--A of the spring body to a fuel rod contacting front side of the plane as described hereinafter. Conversely, the projecting intermediate portions 68 of the outer leaves 70 project rearwardly to the opposite side of the plane A--A, as best illustrated in FIG. 5. Also as indicated upon comparison of FIGS. 3 and 5, end portions 62 have end tabs 74 which project to the forward or fuel rod contacting front side of the plane A--A for overlying the outer surface of a ferrule, as described below. As best illustrated in FIGS. 4 and 5, it will be appreciated that the spring bodies 56a and 56b are disposed in back-to-back relation one to the other. For convenience in handling the spring assembly and assembling the spring bodies and ferrules to form a spacer, the spring bodies 56a and 56b are secured in back-to-back relation, preferably by spot welds 75 along the intermediate portions 68 of outer leaves 70 which abut one another in assembly. In such assembly, the arcuate projections 64 of the central leaves project to opposite sides of the spring assemblies, i.e., toward adjacent fuel rods in adjacent ferrules, as described below. The intermediate portions 68 of the outer leaves 60 lie in back-to-back relation one to the other. With this configuration, the load applied to the springs through the arcuate projections 64 follows a load path through the end portions 62 and through the outer leaves 60 to the back-to-back contact area of projections 68. The central and outer leaves thus act as springs in series, providing a low spring constant, i.e., a soft spring, while affording some flexing in the end portions 62. Particularly note that in the assembly, the plane of each spring body lies essentially parallel to the direction of coolant flow, thus presenting only edges of the spring assembly to the flow. Also, there is only one projection, i.e., the central projection 64 of central leaf 58 which lies in the path of fuel rod insertion into the ferrule. FIG. 8 represents the configuration of the spring assembly when in a stressed condition in use. Referring now to FIGS. 9 and 10, there is illustrated a ferrule 50 according to the present invention. Particularly, ferrule 50 includes a generally cylindrical or tubular element 80 open at its opposite ends and having a rectangular or rectilinear opening or aperture 82 formed through a side wall thereof. As will be appreciated from a review of FIG. 2, the ferrules are secured to one another in a rectilinear array of ferrules, generally a square matrix thereof, and may be coupled one to the other, for example, by welding. In FIG. 10, the spring body 56a is illustrated in position within the opening 82 in bearing engagement against the fuel rod, it being appreciated that, in this form of the invention, only one-half of the spring assembly is illustrated. It will be seen with respect to FIG. 10 that the arcuate projection 64 extends into the ferrule 80 through the rectilinear slot 82 for engagement with the fuel rod 21, biasing it against the stops 54 at the opposite side of the ferrule. The projections 68 lie generally outside the cylindrical confines of the ferrule 80 and the tabs 74 overlie the outer surface of ferrule 80. Referring to FIG. 11, the spring assembly of FIG. 4 is illustrated between adjoining ferrules 80. Thus, as illustrated, the intermediate portions 68 of the outer leaves 60 of the back-to-back disposed spring bodies 56a and 56b are welded one to the other and lie outside of the confines of both ferrules. The spring projections 64, of course, lie within the confines of the ferrules and engage fuel rods 21 within the ferrules 80. The tabs 74 overlie the outer surfaces of the ferrules and locate the spring assembly laterally. The ferrules 80 are secured to one another, preferably by welding. With this spring assembly and ferrule combination disposed in the spacer, for example, as illustrated in FIG. 2, it will be appreciated that the planes A--A of the spring assembly bodies lie generally parallel to the coolant flow so that water droplets impinge on the edges of the spring. The spring bodies are advantageously stamped from sheet metal, while a rectilinear or rectangular slot 82 is formed in the ferrule 80, thus reducing manufacturing costs. Note also that the stiffness of the spring assembly is low because each half of the spring assembly consists of two springs in series. The deflection range is large and, with the exception of the projections 64, the spring assemblies do not have space limiting features. Referring now to the embodiment hereof illustrated in FIGS. 12-18, there is provided a spring formed of sheet metal which acts on a single fuel rod. As illustrated in FIGS. 12-14, the spring includes a spring body 88 having a central leaf 90 joined at opposite ends to end portions 92. A pair of outer leaves 94 are laterally spaced from and straddle the central leaf 90, the outer leaves likewise being connected at their opposite ends to the end portions 92. The central leaf 90 has an arcuate projection 96, preferably medially of the spring, and which projects to a forward side of the spring for engagement against a fuel rod. The spring also includes outer leaves 94 which are projected out of the plane of the spring to a rearward side of the spring as illustrated in FIG. 13. The outer leaves 94 also have a rearwardly extending intermediate, preferably medial projections 98 for bearing against the outer surface of the adjoining ferrule in the spacer. Referring now to FIG. 15, there is illustrated a ferrule 100 comprised of a generally cylindrical element open at its opposite ends and having a generally rectangular or rectilinear opening or aperture 102 through a side wall thereof. Ferrules of this type are arranged in the spacer as illustrated in FIGS. 16 and 18, with the rectangular slot in opposition to a solid wall portion of the adjoining ferrule. As illustrated in FIG. 16, the central portion of the spring 88 lies inwardly of the rectangular slot 102 of ferrule 100, with the medial projection 96 projecting inwardly to engage the fuel rod within the spacer. The end portions 92 bear against the side margins of the rectilinear opening 102. The rearwardly projecting portions 98 of the outer leaves 94 bear against the outer surface of the adjoining ferrule 100. With the ferrules 100 secured to one another, for example, by welding, it will be appreciated that the spring 88, similarly as in the first embodiment, provides a load path through the central leaf 90 to the end portions 92 and through the outer leaves 94 to the contact area between the projections 98 and the outer surface of the adjoining ferrule. Thus, the central and outer leaves act as springs in series, providing a low spring constant. To facilitate insertion of the fuel rods into the ferrules of the spacer, a bar or pin 106 can be inserted between the spring 88 and the ferrule 100, as illustrated in FIG. 17. Thus, the pin 106 may be employed to compress the spring and hence deflect the projections 96 and 98 more towards the central plane of the spring whereby the central portion 96 of the center leaf is spaced back from the position of the fuel rod within the ferrule. By spacing the central spring projection from the fuel rod, the fuel rod may be readily inserted into the spacer, without any force being applied thereto by the spring. A complete fuel rod insertion operation is illustrated in FIG. 18, wherein it will be seen that the pins 106 can be inserted through suitable openings in the peripheral bands surrounding the spacer. Also, it will be noted that the pins 106 compress more than one spring and typically several springs. The pins 106, of course, are withdrawn from the assembly once the fuel rods are inserted into the spacers. While the invention has been described in connection with what is presently considered to be the most practical and preferred embodiment, it is to be understood that the invention is not to be limited to the disclosed embodiment, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims.
050248014	claims	1. A method of updating a core model parameter for a system, comprising the steps of: (a) calculating a calculated value which correlates to a parameter of the model; (b) computing a measured value, which correlates to the parameter, from measured system values; (c) determining a deviation between the calculated value and the measured value; (d) compensating the parameter when the deviation exceeds a predetermined value; and (e) updating the parameter of the model when the parameter is compensated. (a) calibrating the model periodically; (b) measuring enthalpy rise and obtaining neutron detector signal readings, power level and rod position; (c) updating the model using a depletion calculation; (d) calculating analytical values of axial offset and axial pinch; (e) calculating measured actual axial offset and axial pinch values from the enthalpy rise, readings, power level, position and axial offset and pinch expansion coefficients; (f) integrating differences between the measured and analytical values; (g) comparing an absolute value of the differences with predetermined compensation indication limits; (h) adjusting second and third buckling coefficients until the calculated values are compensated, if one of the absolute values is greater than the corresponding limit; (i) updating the parameters when the calculated values are compensated; and (j) displaying the model including the updated parameters. 2. A method as recited in claim 1, wherein the calculated and measured values comprise calculated and measured axial offset values of a nuclear reactor. 3. A method as recited in claim 1, wherein the calculated and measured values comprises calculated and measured values of axial offset and axial pinch of a nuclear reactor. 4. A method as recited in claim 1, wherein the determined deviation is integrated over time and a time integral deviation is compared to the predetermined value to determine when to compensate. 5. A method of updating analytical model parameters of an analytical model used for prediction of distributions in a nuclear reactor core, comprising the steps of: 6. A method as recited in claim 5, wherein step (e) determines axial offset as EQU axial offset=B.sub.1 (DR.sub.t /Q)+B.sub.2 (DR.sub.b /Q)+B.sub.3 bp 7. A method as recited in claim 6, further comprising updating boron concentration.
047724463	abstract	Apparatus for releasably engaging an elongated member that is releasably connected to the top nozzle of a nuclear reactor fuel assembly. The top nozzle has an adapter plate disposed at its lower end having at least one passageway therethrough through which the elongated member is disposed. The elongated member has a releasable latching structure at one end that is able to be engaged by the apparatus, having at least one latching member movable between a latched position in which the latching member is able to engage said adapter plate and secure the absorber rod in a stationary relationship with respect to the adapter plate and an unlatched position in which the latching member is able to disengage from said adapter plate so that the elongated member be removed from the fuel assembly. The apparatus comprises a hollow releasing member for moving the latching member of the latching structure between its latched position and its unlatched position, an engaging member connected to the releasing member and extending downwardly through the hollow portion of the releasing member, and an actuating member extending downwardly through the hollow portion of the releasing member and coacting with the engaging member to releasably engage the latching structure.
	abstract	A rotatable fixture for use in a vibration or seismic test is capable of shortening the time for dismantling the fixture. The rotatable fixture of the present invention is aimed to obviate the non-universalness of the conventional fixture and remove the drawbacks that the conventional fixture has to be dismantled before rotating to a certain angle for test. The rotatable fixture of the present invention includes a base, a rotating disc, a fastening frame, and fixing plates. The rotatable fixture has a basic natural frequency which is distant from the test frequency band. Meanwhile, the drawbacks that the fixture has to be dismantled during the test process is eliminated by the special connection between the base and the rotating disc. Moreover, due to the versatile design of the fastening frame, the rotatable fixture can allow a variety of devices under test to be secured thereto to attain universalness.
	abstract	A transmitter device includes a neutron detector structured to detect neutron flux, a capacitor electrically connected in parallel with the neutron detector, a gas discharge tube having an input end and an output end, and an antenna electrically connected to the output end. The input end is electrically connected with the capacitor. The antenna is structured to emit a signal corresponding to the neutron flux.
054328284	summary	The invention relates to a method of replacing an adaptor for penetrating the vessel head of a nuclear reactor cooled by pressurized water. Pressurized-water nuclear reactors generally include a vessel enclosing the core of the reactor which is submerged in pressurized water for cooling the reactor. The reactor vessel of cylindrical general shape includes a head of hemispherical shape which can be attached to its upper portion. The head is pierced with openings at each of which is fixed a tubular penetration part constituting an adaptor providing the passage and the actuation mechanism for moving a follower of a rod for controlling the reactivity of the core or a penetration passage for a means of measurement inside the core, such as a thermocouple column assembly. The tubular part is generally shrink-fitted inside the opening of the head and is fixed in a sealed manner by means of a weld, deposited in a bevel machined on the concave inner surface of the vessel head, around the penetration opening for the adaptor. To the end portion of each of the adaptors is fixed a mechanism for actuating the movement of a rod for controlling the reactivity of the core or a means for supporting a thermocouple column assembly. Inside each of the adaptors for penetrating the vessel head providing the passage of a follower of a reactivity control rod is fixed, in a position coaxial in relation to the tubular adaptor and with a certain radial clearance, a thermal sleeve which includes a diametrically widened portion resting on a bearing surface located at the upper portion of the bore of the adaptor. In order to increase the operational reliability and safety of nuclear reactors and to extend the lifetime of these reactors, the operators are obliged to carry out an increasing number of inspections on the various elements constituting the nuclear reactor. For example, the state of the parts for penetrating the vessel head is inspected in order to be sure of the integrity of these parts, after a certain time of operating the reactor, in particular in the area where these tubular parts are welded to the vessel head. These inspections and repairs are carried out during a shutdown of the nuclear reactor, the vessel head being dismounted and placed on an inspection and repair stand. The head is placed on the inspection and repair stand in a position corresponding to its surface position, the convex external surface of the head facing upwards and the concave internal surface facing downwards. The presence of defects in the adaptors, which may, in some cases, be constituted by cracks, can also be detected during hot pressure testing of the nuclear-reactor vessel. In some exceptional cases, defects constituted by cracks penetrating the wall of the adaptor may be present. In the case where defects are detected, it is necessary to carry out a repair, on the inspection and repair stand, before replacing the head in the service position on the nuclear-reactor vessel. In some cases, in particular in the very rare case where a crack of an adaptor would be penetrating, it is necessary either to replace the vessel head completely or to replace the defective adaptor by a new adaptor. Obviously the complete replacement of the head is an extremely expensive operation, one which requires having a complete head in reserve on the nuclear-reactor site. The operation of replacing a defective adaptor by a new adaptor, which requires removal by machining of the lower portion of the adaptor to be replaced and of the weld for fixing the head, as well as welding of the replacement adaptor, is lengthy and tricky to perform, in particular on account of the fact that the internal surface of the head is highly radioactive. Such a repair, which requires the installation of a protective structure, can turn out to be lengthy and expensive. The object of the invention is therefore to provide a method of replacing an adaptor for penetrating the vessel head of a nuclear reactor, of rounded shape, disposed, after dismounting, so as to have a convex outer surface facing upwards and a concave inner surface facing downwards, each of the adaptors having a vertical disposition and being fixed in a penetration opening in the head by means of a weld in the vicinity of the concave inner surface and carrying an item of equipment resting on one end of the adaptor, this end projecting outside the head, the method consisting in machining a lower portion of the adaptor to be replaced, in the vicinity of the inner surface of the head, and at least one portion of the weld for fixing the adaptor, in extracting the adaptor to be replaced from the opening in the head, in inserting a replacement adaptor into the opening and fixing it therein, and in welding the replacement adaptor to the head, this method being able to be implemented in a simple manner and enabling the cost of the replacement operation to be reduced. With this object in mind, the replacement is performed from a repair area located above the convex face of the head and at least one of the operations of machining the adaptor to be replaced after dismounting the item of equipment resting on its upper end and/or of welding the replacement adaptor is carried out by inserting an appropriate tool via the upper end of the adaptor.
	abstract	A method and system for imaging an object are described herein. A scatter image of the object is generated at a projection angle. In generating the scatter image, a non-grid image of the object is acquired using a radiation source and a detector. An aperture plate is positioned between the object and the detector and a first grid image of the object is acquired. The aperture plate includes a plurality of apertures positioned on a grid. The aperture plate is moved to a second position and a second grid image of the object is acquired. A scatter image of the object is generated based on the non-grid image, the first grid image, and the second grid image and stored.
	claims	1. A method of manufacturing a scintillator panel configured to convert radiation into scintillation light, the method comprising:a first process of forming a plurality of convex sections that protrude in a predetermined direction from a rear surface toward a front surface and concave section defined by the convex sections, on the front surface of the substrate having the front surface and the rear surface;a second process of forming first scintillator units respectively extending from the convex sections of the substrate in the predetermined direction through crystal growth of a columnar crystal of a scintillator material; anda third process of radiating a laser beam to contact portions of the first scintillator units extending from the adjacent convex sections and separating the first scintillator units extending from the adjacent convex sections by scanning the concave section with the laser beam. 2. The method of manufacturing the scintillator panel according to claim 1, further comprising, before the third process, a fourth process of forming second scintillator unit on bottom surfaces of the concave section of the substrate. 3. The method of manufacturing the scintillator panel according to claim 2, wherein,in the first process, the concave section defined in a lattice shape are formed on the front surface of the substrate by forming the convex sections to be arranged on the front surface of the substrate in a two-dimensional array, andin the fourth process, a thickness of the second scintillator unit in a crossing region of the concave section is larger than a thickness of the second scintillator unit at a position other than the crossing region. 4. A scintillator panel configured to convert radiation into scintillation light, the scintillator panel comprising:a substrate having a front surface and rear surface, the substrate is formed with a plurality of convex sections protruding from the front surface in a predetermined direction from the rear surface toward the front surface, and concave section defined by the convex sections; anda plurality of first scintillator units respectively extending from the convex sections in the predetermined direction and separated from each other,wherein the first scintillator units are respectively formed through crystal growth of a plurality of columnar crystals on the convex sections, andat least a part of the columnar crystals constituting the first scintillator unit over bottom surface of the concave section are fused and adhered to each other through radiation of a laser beam. 5. The scintillator panel according to claim 4, further comprising second scintillator unit formed on the bottom surfaces of the concave section of the substrate. 6. The scintillator panel according to claim 5, wherein the convex sections are arranged on the front surface of the substrate in a two-dimensional array,the concave section is defined on the front surface of the substrate by the convex sections in a lattice shape, anda thickness of the second scintillator unit in a crossing region of the concave section is larger than a thickness of the second scintillator unit at a position other than the crossing region. 7. The scintillator panel according to claim 5, wherein at least a part of the second scintillator unit is solidified after being melted by radiation of the laser beam. 8. A radiation detector configured to detect radiation, the radiation detector comprising:a substrate comprising a plurality of photoelectric conversion elements and having a front surface and rear surface, the substrate is formed with a plurality of convex sections protruding from the front surface in a predetermined direction from the rear surface toward the front surface, and concave section defined by the convex sections, the plurality of convex sections being formed so that the plurality of convex sections respectively correspond to the plurality of photoelectric conversion elements; anda plurality of first scintillator units respectively extending from the convex sections in the predetermined direction and separated from each other,wherein the first scintillator units are respectively formed through crystal growth of a plurality of columnar crystals on the convex sections, andat least a part of the columnar crystals constituting the first scintillator unit over bottom surface of the concave section are fused and adhered to each other through radiation of a laser beam. 9. The scintillator panel according to claim 6, wherein at least a part of the second scintillator unit is solidified after being melted by radiation of the laser beam. 10. The radiation detector according to claim 8, further comprising second scintillator unit formed on the bottom surfaces of the concave section of the substrate. 11. The radiation detector according to claim 10, wherein the convex sections are arranged on the front surface of the substrate in a two-dimensional array,the concave section is defined on the front surface of the substrate by the convex sections in a lattice shape, anda thickness of the second scintillator unit in a crossing region of the concave section is larger than a thickness of the second scintillator unit at a position other than the crossing region. 12. The radiation detector according to claim 10, wherein at least a part of the second scintillator unit is solidified after being melted by radiation of the laser beam. 13. The radiation detector according to claim 11, wherein at least a part of the second scintillator unit is solidified after being melted by radiation of the laser beam. 14. A method of manufacturing a radiation detector configured to detect radiation, the method comprising:a first process of forming a plurality of convex sections that protrude in a predetermined direction from a rear surface toward a front surface and concave section defined by the convex sections, on the front surface of the substrate comprising a plurality of photoelectric conversion elements and having the front surface and the rear surface, the plurality of convex sections being formed so that the plurality of convex sections respectively correspond to the plurality of photoelectric conversion elements;a second process of forming first scintillator units respectively extending from the convex sections of the substrate in the predetermined direction through crystal growth of a columnar crystal of a scintillator material; anda third process of radiating a laser beam to contact portions of the first scintillator units extending from the adjacent convex sections and separating the first scintillator units extending from the adjacent convex sections by scanning the concave section with the laser beam. 15. The method of manufacturing the radiation detector according to claim 14, further comprising, before the third process, a fourth process of forming second scintillator unit on bottom surfaces of the concave section of the substrate. 16. The method of manufacturing the radiation detector according to claim 15, wherein,in the first process, the concave section defined in a lattice shape are formed on the front surface of the substrate by forming the convex sections to be arranged on the front surface of the substrate in a two-dimensional array, andin the fourth process, a thickness of the second scintillator unit in a crossing region of the concave section is larger than a thickness of the second scintillator unit at a position other than the crossing region.
	summary
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	description	This application claims priority to U.S. Provisional Application Ser. No. 62/369,443 filed on Aug. 1, 2016, which is incorporated in its entirety. The field of the disclosure relates generally to HEPA filter airflow distribution systems, and more particularly to such systems for isolator working volumes or radionuclide generators. Shielded nuclear containment chambers, referred to as hot cells or isolators, are used in the radiopharmaceutical and nuclear energy industries to protect personnel from the radioactive material contained therein and radiation emitted therefrom. As used herein, hot cells and isolators are collectively referred to as isolators. In the radiopharmaceutical industry, radioactive material used for diagnostic and therapeutic purposes may be produced using a radionuclide generator, such as column housed in an isolator. Additionally, associated radionuclide process equipment is typically housed in an isolator. Air filters such as HEPA filters are used to provide pharmaceutical grade sterile filtered air to isolators, but current systems are not completely satisfactory. Accordingly, an improved system for HEPA filter airflow distribution is needed. This Background section is intended to introduce the reader to various aspects of art that may be related to various embodiments and aspects of the present disclosure, which are described and/or claimed below. This discussion is believed to be helpful in providing the reader with background information to facilitate a better understanding of the various embodiments and aspects of the present disclosure. Accordingly, it should be understood that these statements are to be read in this light, and not as admissions of prior art. In one aspect, an air distribution system for supplying filtered air to an isolator working volume comprises an inlet including a HEPA filter, and an outlet including a slidably mounted sintered panel. The slidably mounted panel is positioned between the HEPA filter and the isolator working volume. In other aspects, a method for supplying filtered air to an isolator working volume comprises filtering the air through a HEPA filter into a plenum chamber defined by the volume between (i) the outlet surface of the a HEPA filter, (ii) the inlet surface of a slidably mounted sintered panel, and (iii) the isolator walls. The slidably mounted sintered panel is positioned between the HEPA filter and the working volume of the isolator. The air pressure in the plenum chamber is greater than the air pressure in the isolator working volume. In some other aspects of the disclosure, an air distribution system for supplying filtered air to an isolator working volume is provided. The system comprises: a HEPA filter for filtering air to the isolator; a track mounted between the HEPA filter and the working volume, and a sintered panel slidably mounted on the track; and a plenum defined by the volume between the HEPA filter outlet surface, the sintered panel inlet surface, and the isolator walls. The sintered panel being slidable between a first closed position directly under the HEPA filter, and a second open position enabling access to the HEPA filter. Various refinements exist of the features noted in relation to the above-mentioned aspects. Further features may also be incorporated in the above-mentioned aspects as well. These refinements and additional features may exist individually or in any combination. For instance, various features discussed below in relation to any of the illustrated embodiments may be incorporated into any of the above-described aspects, alone or in any combination. Corresponding reference characters indicate corresponding parts throughout the Figures. Isolators are used to provide separation between a processing environment (e.g., an isolator working volume) and the environment external to the processing environment. Isolators typically comprise (i) gas (e.g. air) supply filtration and/or process gas (e.g., air) discharge filtration and (ii) a working volume containing process equipment or in-process materials. In some aspects, process gas may contain entrained liquid and/or solid particulate. In some aspects, the isolator prevents or mitigates contamination of the external environment with hazardous components from the working volume, such as radioactive material, chemical toxins and/or biological toxins. In some other aspects, the isolator prevents or mitigates contamination of the working volume with contaminants from the environment external to the working volume, such as microbes that could compromise the sterility of material in the working volume. Radioactive material is used in nuclear medicine for diagnostic and therapeutic purposes by injecting a patient with a small dose of the radioactive material, which concentrates in certain organs or regions of the patient. Radioactive materials typically used for nuclear medicine include Technetium-99m, Indium-111, Thallium-201, and Strontium-89, among others. Such radioactive materials may be produced using a radionuclide generator manufactured in a shielded nuclear containment chamber, referred to as a hot cell or isolator. Radionuclide generators generally include a column that has media for retaining a long-lived parent radionuclide that spontaneously decays into a daughter radionuclide that has a relatively short half-life. The column may be incorporated into a column assembly that has a needle-like outlet port that receives an evacuated vial to draw saline or other eluent liquid, provided to a needle-like inlet port, through a flow path of the column assembly, including the column itself. This liquid may elute and deliver daughter radionuclide from the column and to the evacuated vial for subsequent use in nuclear medical imaging applications, among other uses. Additionally, associated radionuclide process equipment is typically housed in the hot cell or isolator. Hot cells or isolators, are used in the radiopharmaceutical and nuclear energy industries to shield and protect personnel from the radioactive material contained therein and radiation emitted therefrom. As used herein, hot cells and isolators are collectively referred to as isolators. Suitable shielding materials from which isolators may be constructed include, for example and without limitation, lead, depleted uranium, and tungsten. In some embodiments, isolators are constructed of steel-clad lead walls forming a cuboid or rectangular prism. In some embodiments, an isolator may include a viewing window constructed of a transparent shielding material. Suitable materials from which viewing windows may be constructed include, for example and without limitation, lead glass. Air filters such as HEPA filters are used to provide pharmaceutical grade sterile filtered air to isolators. For instance, four HEPA filters may be used in a 2×2 matrix or six HEPA filters may be used in a 2×3 matrix. In some embodiments, air supply HEPA filters are positioned in the ceiling portion of the isolator and provide for a generally uniform and unidirectional flow of filtered air over a manufacturing area of the isolator. In accordance with the present disclosure, the isolator air supply further comprises a diffusion grid or membrane positioned between the outlet side of the HEPA filters and the isolator working volume thereby forming a plenum chamber that provides for improved air distribution as compared to HEPA filters alone. The diffusion grid or membrane is mounted such that it is allowed to slide to provide HEPA filter access. Isolators often use a diffusion grid or membrane situated between the HEPA filters and the isolator working volume in order to create a full ceiling HEPA filtered unidirectional airflow. Such diffusion grids or membranes must be manually removed in order to access HEPA filters and perform HEPA filter certification and then reinstalled after certification is complete. Problematically, manually handling diffusion grids or membranes can cause damage to filters and the grids/membranes, and thereby increase testing time and thus process cycle time. Further, because of radiation in the isolator, manually removing diffusion grids may pose a safety risk. FIG. 1 is a side view of an isolator 100 air distribution system 1 of the present disclosure. As shown, the air distribution system 1 generally comprises one or more HEPA filters 10 for filtering isolator supply air 5, a sliding panel grid 20 having sintered panels, where the HEPA filter 10 outlet surface, the sintered panel 20 inlet surface, and the isolator 100 walls 101, 102 and 103 define a plenum area or volume 30. The HEPA filters are each mounted in a frame (not depicted) and define the plenum inlet 35. The sliding panel grid 20 is mounted in a frame (not depicted), the frame comprising tracks affixed to isolator side walls for mounting the sintered panels, and the panels are slidable on the tracks. The sintered panels define the plenum outlet 36. As shown in FIG. 1, the plenum 30 is positioned above a manufacturing area 50 having a telemanipulator port 40 that allows an operator outside the hot cell environment (i.e., within the surrounding room or lab) to manipulate and transfer column assemblies within the isolator 100 environment with an automated, semi-automated or manually operated telemanipulator (not depicted) through the telemanipulator port. The isolator 100 may also have one or more ports for barcode scanner equipment, termed “barcode periscopes” (not depicted). Each radionuclide generator is labeled with a unique 2-dimensional barcode and the barcode is read prior to certain process steps in order to assure generator process control and traceability. Barcode scanner ports are positioned below the air distribution system 1. FIG. 2 is a top view of is a top view of an isolator 100 HEPA filter housing assembly of the present disclosure depicting four HEPA filters 10 arranged in a 2×2 matrix. FIG. 2 is not limiting and other HEPA filter arrangements are within the scope of the present disclosure, for instance, one HEPA filter, two HEPA filters, and six HEPA filters arranged in a 2×3 matrix. FIG. 3 is a perspective view of an air distribution system 1 of the present disclosure in a maintenance mode wherein one or more HEPA filters may be accessed from a hot cell door (not shown). HEPA filters 10 are suitably mounted in a frame (not shown) for filtering isolator 100 supply air. Two HEPA filters of a 2×2 matrix of four HEPA filters are depicted, but other arrangements of HEPA filters are within the scope of the present disclosure, for instance, one, two or six HEPA filters. As shown, the air distribution system includes a sliding panel grid 20 and formed from Panel 60 and Panel 61. Other arrangements of panels are within the scope of the present disclosure, including for example one slidable panel, three slidable panels, and four slidable panels. Each of Panel 60 and Panel 61 is mounted in a track 50, and the panels are slidable on the track. The volume defined by the HEPA filters 10 outlet surface, the inlet surface of the sliding panel grid 20, and the isolator side walls define a plenum chamber or plenum volume. The HEPA filters 10 define the plenum inlet and the sliding panel grid 20 defines the plenum outlet. The plenum chamber or plenum volume is positioned above a manufacturing area having a telemanipulator port 40 as described elsewhere herein. In any of the various aspects of the present disclosure, the sintered panels are mounted in tracks 50, and the panels are slidable on the tracks 50. In some aspects of the disclosure, as depicted in FIG. 3, the panels comprise slides 70 that are slidable on tracks 50. Any slide mount (i) suitable to allow diffuser panel movement with an external device and (ii) capable of forming a seal sufficient to maintain a higher, or positive, pressure in the plenum relative to the isolator manufacturing area is within the scope of the present disclosure. In some aspects, stainless steel inverted V groove slides on plastic runners may be used. In embodiments including two or more panels, the tracks are positioned at different elevations within the isolator in order to allow the panels to slide under or over an adjacent panel such as depicted in FIG. 3. The slides of each panel are engaged with the track, and move along the track, thereby allowing for axial movement of each panel. As previously described, the air distribution systems of the present disclosure include a plenum defined between the HEPA filters and the sintered panels. Supply air is filtered through the HEPA filter wherein the HEPA filter outlet surface defines the plenum inlet and wherein the sintered panel inlet surface defines the plenum outlet. HEPA filters are known in the art and generally include a rigid or semi-rigid frame and a sealed filter medium made of a continuous sheet of close spaced pleated media of randomly arranged fibers. One example of a suitable fiber is fiberglass. HEPA filters for radionuclide applications generally comply with ASME AG-1, FC-3000 or FK-3000. Suitable media provide for a maximum air velocity of 5.0 feet per minute and include folded filter media with corrugated separator and supports (Type A), minipleat media with glass ribbon or non-combustible thread separators (Type B), continuous corrugated filter media folded without separators (Type C), and filters that include glass or non-combustible thread separators (Type D). Sintered panels are known in the art and may be constructed from sintered metal. In some aspects, the metal is stainless steel. In general, sintered metal membranes are formed from compressed and heated metal granules, powders or fibers thereby fusing the metal, in the essential absence of melting, in a porous network. Sintered air filters provide for uniform porosity, selectivity and precision, thereby creating uniform pressure drop and airflow across the filter face. FIG. 4 is a perspective view of an isolator 100 air distribution system 1 of the present disclosure in a manufacturing mode where the sintered panels are in a closed position. Panel edge 61B of Panel 61 is sealably engaged with the isolator wall 1 80 and Panel edge 60A of Panel 60 is positioned such that the indicated Panel 60 edge 60A is sealably engaged against the isolator wall 2 81. Panel 60 and Panel 61 overlap and seal against one another. In some aspects, a seal between Panels A and B and between the panels and the isolator walls is achieved with silicone sealing strips, among other suitable materials. Isolator manufacturing area pressure may be controlled at a negative or positive pressure relative to the surrounding environment and/or relative to adjacent process equipment. In some embodiments, the isolator manufacturing area pressure is controlled at negative pressure, such as from about −0.01 inches H2O to about −0.15 inH2O. The plenum of the present disclosure operates at a higher, or positive, pressure relative to the isolator manufacturing area. The positive pressure zone in the plenum provides for an even airflow distribution over the entire surface area of the hot cell while also providing a protective barrier for the filters within the cell. Further, improved air distribution in the plenum created by the sintered panel grid allows for a reduced number of filters while still allowing full dispersion of clean air flow. In a maintenance mode, the sintered panels may be positioned as indicated in FIG. 3 to allow access to the exposed row of HEPA filters. The panels may be positioned opposite to that depicted in FIG. 3 such that both panels are located proximate to an isolator wall thereby allowing access to the row of HEPA filters located above, and obscured by, Panel 60 and Panel 61 of FIG. 3. The slidable panels provide for full access to HEPA filters for integrity testing, repair or replacement without the need to remove grids, membranes, telemanipulators and/or barcode periscopes from the isolator. The present disclosure includes methods of supplying a generally uniform and unidirectional flow of filtered air to an isolator, such as the hot cell of a radionuclide generator, using a plenum chamber defined by a volume between the outlet surface of one or more plenum inlet HEPA filters and the inlet surface of one or more plenum outlet sintered panels as described elsewhere herein, wherein the sintered panels are positioned between the HEPA filters and the hot cell working volume. The supply air is forced through the one or more HEPA filters and into the plenum chamber. The air pressure in the plenum chamber is greater than the air pressure in working volume of the hot cell. The filtered air passes through the sintered panels in an evenly distributed air flow over the entire hot cell working volume. The sintered panels may be moved manually or automatically. For instance, the panels may be moved manually with an external device. In some embodiments, the panels may be operably connected to an automated mover. Examples of automated movers include a pneumatically actuated cylinder, such as a rodless cylinder as known in the art, and a motor-driven ball-screw apparatus as known in the art. When introducing elements of the present invention or the embodiment(s) thereof, the articles “a”, “an”, “the” and “said” are intended to mean that there are one or more of the elements. The terms “comprising”, “including” and “having” are intended to be inclusive and mean that there may be additional elements other than the listed elements. This written description uses examples to disclose the invention, including the best mode, and also to enable any person skilled in the art to practice the invention, including making and using any devices or systems and performing any incorporated methods. The patentable scope of the invention is defined by the claims, and may include other examples that occur to those skilled in the art. Such other examples are intended to be within the scope of the claims if they have structural elements that do not differ from the literal language of the claims, or if they include equivalent structural elements with insubstantial differences from the literal languages of the claims.
046706585	description	DESCRIPTION OF THE PERFERRED EMBODIMENTS In FIG. 1 the sheet 10 includes a radio-opaque ply 14 formed of a support matrix and a radio-opaque compound supported by the matrix. A thin, impermeable to liquid polyethylene base ply 12 is attached to and positioned below the radio-opaque ply 14. An absorbant surgical drape ply 16 is attached to and possitioned above radio-opaque ply 14. Plies 12 and 16 reduce any possibility of leakage of the radio-opaque compound into the surgical field. Additionally, ply 16 acts as an absorbant layer to sponge-up blood and other fluids. The radio-opaque ply 14 is a gauze sheet which carries barium sulfate powder. The gauze sheet can be impregnated with the barium sulfate by soaking the gauze sheet in barium sulfate suspension and then evaporating the water thus providing an even distribution of the barium sulfate in the ply 14. Alternatively it might be possible to apply the barium sulfate dry to the gauze. The amount of powdered barium sulfate in the gauze ply can be anywhere from 350 to 730 grams per square foot. This will block anywhere from fifty percent (50%) to one hundred percent (100%) of the incident radiation up to the X-ray frequency generated at a 120 KV setting on the radiology machine. The plies 14 and 16 can be anywhere from one-quarter to one-half inch thick each and the ply 12 can be in the order of two to four mils (0.002 to 0.004 inches). The following test was conducted on the effectiveness of the radio-opaque ply. A fluoroscopic exposure was made of an upper abdomen phantom using a ten inch field, at an X-ray machine setting of ninety KV and three milli-amperes. A pocket radiation dosimeter was placed in the left inguinal region (groin) supported by three-fourths of an inch (0.75 in) thick qauze and held in place by tape. The phantom was then fluoroscoped for five minutes. The radiation dose registered on the pocket dosimeter was eleven millirads. The procedure was repeated replacing the gauze with a three and one-half inch by three inch barium sulfate soaked sheet containing approximately fifty-five (55) grams of barium sulfate thus proving a sheet having a density of about 755 grams per square foot. The sheet measured approximately 0.75 inch in thickness. The phantom was again fluoroscoped for five minutes. The dose registered on the pocket dosimeter was less than one millirad, a reduction of greater than ten to one. Because of the cumulative effect of radiation, any reduction in radiation, even by a small amount, is a benefit to the individual exposed and thus is desirable. Accordingly, this reduction to one-tenth of the exposure that otherwise would occur reduces the risk to all operating personnel. It is contemplated that one useful size of this protective sheet will be in the range of one square foot to two square feet. Sheets of that size can be placed under a doctor's hands during an operative procedure. Such sheets might also be placed adjacent to each other in a pattern with their edges overlapped to define an area in which radiation is not blocked and to provide a blockage of radiation outside of that area. In the case of sheets of this size, the various plys 12, 14, 16 could be held together along their edges by, for example, surgical sterile adhesive tape or by thermal bonding or even by stitching. Where larger sheets may be employed, it might be desirable to include strips of double faced adhesive between the plys to facilitate the handling of the sheet. In the FIG. 2 embodiment, the sheet 20 is a three ply sheet having a center ply 24 that is substantially the same as is the radio-opaque ply 14 of the sheet 10 shown in FIG. 1. However, in this FIG. 2 embodiment, the radio-opaque ply 24 is sandwiched between two liquid impermeable polyethylene plies 22, each of which plies are substantially similiar to the ply 12 in the FIG. 1 embodiment. This FIG. 2 embodiment might be used in circumstances where the liquid absorption characteristic is not required. For example, the FIG. 2 embodiment might be used as a partition around a patient to provide protection for personnel required in the operating room from scatter radiation. However the sheet 20 can be adapted to be used in place of an actual surgical drape by slipping the sheet 20 into an envelope formed of two sheets 32 of normal surgical drape material as shown in FIG. 3. The result is a sheet 30 that is composed of five plies and which provides the liquid absorbant feature. FIG. 4 shows a forth embodiment of a sheet 40 in which powered barium sulfate 42 is carried in plastic pockets 44 of bubble packet type material. A sheet of this barium sulfate filled bubble pack material 45 is attached to a ply of surgical drape material 46 similar to the ply 16 in the FIG. 1 embodiment. In the FIG. 4 embodiment as shown, the bubble pack sheet 45 is one in which a flat ply 45a forms the upper surface of the pockets 44 and a formed sheet 45b is attached to the flat sheet 45a to define individual pockets 44. Certain applications of the sheets 10, 20, 30, 40 of this invention may require much larger sheet material than the one to two square foot material mentioned above. Large protective sheets can be laid over a patient or under a patient or both during an operating procedure. In such a case, the large sheet will require an opening therethrough so that the primary beam of x-ray will not be blocked. Although the above description emphasizes the value of using these sheets to block secondary radiation from impinging on operating personnel, it should be recognized these sheets can be used to protect genitalia, or other body parts, of patients during radiographic procedures. It should be recognized that the use of large size sheets of this invention over and/or below a patient during a fluoroscopic surgical procedure serves to prevent the scatter radiation from passing through the sheets and thus tends to protect all personnel in the operating room. In many fluorscopic procedures, such as angiographic procedures, a catheter is manipulated and introduced through the femoral artery. The tip of the catheter is monitered under fluoroscopic control. If a protective sheet made in accordance with the present invention is used in the inguinal area, under or over the operator's hands, which is the region where the catheter is introduced, a decreased radiation dosage to the operator's hands by a factor of ten would result. Because of the relatively low cost of the protective sheets of this invention, they are disposable and thus need not be used except for a single operation. Thus, sterility in the operating zone is enhanced and cross contamination is avoided. One advantage of maintaining sterility in the operating zone is that a doctor's movements are less inhibited in that his or her contact with the sterilized protective sheet of this invention does not destroy the sterilization procedures.
056051719	description	DESCRIPTION OF THE PREFERRED EMBODIMENT Bulk silicon, because of its indirect band-gap, does not normally display optical excitation. However, a microporous form of crystalline silicon can be fabricated via electrodissolution of regular bulk silicon in HF solution as taught in U.S. Pat. Nos. 5,272,355; 5,301,204; and 5,285,078, the disclosures of which are herein incorporated by reference. Very high porosity (from about 50 to over about 90 percent by volume) can be generated. It was discovered recently that porous silicon (PS) is optically active. It exhibits photoluminescence as well as electroluminescence in the visible region at room temperature. This discovery opens up the possibility of being able to utilize silicon, the most fundamental of electronic materials, in an optoelectronics context. In porous silicon the pore sizes are practically microscopic (as small as a few nm) preferably in the range of from about 1 to about 100 nanometers. The silicon crystallites that make up the walls of the pores are of similar microscopic dimension. Despite extensive work devoted to the subject in recent times there does not yet exist a generally accepted understanding of the mechanism and the process of porous silicon formation. However, certain general features of the process are known. The infiltrating attachment on the silicon at the silicon-electrolyte interface results in very convoluted interconnected pore structures with pore tips as small as only a few nm across. This means that during the dissolution process the pore tips are in an active state, whereas the pore surfaces behind the pore tip regions remain passive. So the pore has continuous forward growth in the tip region while there is much slower "sideways" growth in the pore region behind the pore tips. This results in the formation of long but narrow pores characteristic of many types of porous silicon. The aspect ratio can be as much as 1 to 10 or even 1 to 100. There are at present a number of suggestions on the mechanism responsible for the conversion between active and passive states. They range from curvature-induced effects (due to the proximity to the pore tips) to quantum confinement (due to the small sizes of the crystallites making up the pore tips) to quantum confinement (due to the small sizes of the crystallites making up the pore walls) and to the presence of depletion layers frequently found near semiconductor/electrolyte interfaces. Curvature-induced effects may arise from the fact that the pore surface in the region around the tip is necessarily nonplanar. In atomistic terms this necessitates a higher density of steps in the vicinity of that surface region. The increase in abundance of these surface line defects and their associated point defects such as kinks may in turn give rise to higher surface reactivity. However, these mechanism need not be mutually exclusive. They could be in effect concurrently, leading to the simultaneous evolution of more than one type of microstructural feature. Nevertheless, independent of the physical origin of the optical properties of PS, it is known that hydrogenic species exist on the pore surfaces of this material. The present invention exploits this fact by depositing tritium species onto those same interior pore surfaces via the same electrodissolution process that has been employed to synthesize porous silicon except for one key modification. The electrolyte is composed of a mixed hydrogen fluoride (HF)/tritiated fluoride (TF) solution. The relative HF/TF composition determines the relative amount of tritium that is deposited as surface chemical species (e.g., as Si--H--T or Si--O--H) onto the interior pore surfaces in the porous silicon network. That is the amount of tritium deposited depends on the ratio of HF to TF in the electrolyte solution. An alternative synthesis route of making tritiated porous silicon (TPS) is to equilibrate conventional PS with a vapor containing HT/T.sub.2 gas in a diluent such as H.sub.e or H.sub.2. This process depends on chemical vapor infiltration and exchange reaction between the tritium in the gas phase with the hydrogen present in the pore surfaces of PS. The amount of tritium deposited is controlled by the ratio of tritium present in the diluent gas. This is a vapor deposition method unlike that of the electrolytic route which depends on liquid phase processes. Both methods will deposit tritium onto the interior pore surfaces of the porous silicon resulting in a new material TPS. In this application the chemisorbed species is desired for thermal stability of the device with the tritiated porous silicon (TPS) acting as the luminescent material. For the tritiated porous silicon samples, luminescence spectra is to be collected sing a spectrophotometer. Whereas a deuterated silicon photoluminescence (PL) data is ascertained in tritiated silicon radioluminescence (RL) measurements are made. For both PL and RL the latter stages of the process are the same, namely electron-hole excitation and subsequent radiative recombination. This means that for the experimental results the major difference lies in the nature of the excitation source. In PL the excitation source is photons (i.e., light). In RL the corresponding source is electrons. However, the optical measurements are closely similar aside from the nature of the excitation source. It is important to note that even though in TPS one uses electrons as the excitation source the process has important differences with electroluminescence (EL). In EL the free charge carriers (such as electrons or holes) are injected into the the materials and directed towards the region of luminescent centers under the action of an externally applied field. These energetic charge carriers dissipate their energy in the host material via mainly electron-hole pair excitations and photon generations. An example is the commonly available LED (light emitting diode) in which a p-n junction type device is utilized under a forward bias that provides the external field. When EL efficiency is low, the cause has often been attributed to the difficulty of injecting carriers into the system. This can arise for a wide variety of reasons, such as high contact resistance. In fact the best EL performance from conventional porous silicon at present has been obtained by using liquid electrolyte contact. Wet contact is provided by immersing the PS layer in a liquid electrolyte which provides the medium for carrier injection. For p-type silicon substrate the carriers injected from the electrolyte would be the minority type (i.e., electrons). For technological applications there would be obvious difficulties in attempting to use PS-based devices that have to depend on liquid phase contacts. Realistic applications would require either solid state contact or some form of contactless designs, as has been suggested. The main idea behind TPS minimizes the difficulty of carrier injection due to two factors. Firstly, the electron emission source (namely the tritium) is dispersed microscopically throughout the host material. Secondly, the range of the electrons (of the order of a fraction of a micron) implies that the carrier injection would not be an issue. Referring to FIG. 1, TPS 10 can act as a stand alone power source emitting light 11. Its function is similar to other advanced stand-alone energy generating/storage devices such as batteries. However, unlike advanced batteries, which involve power levels in the range of KW to MW (particularly for utilities and transportation applications), the operating range of TPS is expected to be mWs or less. It is suited for the low-level power requirement characteristic of highly integrated devices in optoelectronics and control technology. Strictly speaking, TPS is not an energy storage device like a battery. Once in place TPS would generate light energy continuously. However, the very long time constant (11-12 years) allows TPS to operate as an energy generating source, while also effectively functioning like an energy storage device in the sense that power is available "on call" without the need for an external source of energy. Another important application of this invention is to utilize TPS as a self-powered illumination source for low-level lighting applications where stand-alone and low or zero maintenance capabilities are of the highest priority. Another important use of the invention involves the direct coupling of TPS to photovoltaic cells 15 to generate electricity. These uses illustrate the different ways that TPS can be utilized in energy technology applications. The photovoltaic application involves the direct conversion of luminescent energy to electrical power. The lighting technology uses TPS as the illumination source. The optoelectronic application utilizes the optical output 11 from TPS 10 as a power supply and also as an interconnect between different electronic components 20. The common denominators to all these options are the stand-alone capability, the compactness of an inorganic solid state source, and the practically uninterruptable nature of its operating characteristics. These factors make TPS well-suited for energy technology applications that involve remote site operation, industrial process monitoring and control in limited access environments, back-up retention of information/data, operation of energy devices (such as exit signs), and instrumentation panels in case of main power disruption. Further, TPS can be used as an on-board stand alone power supply for a silicon micromachine technology. The fact that silicon has excellent electronic characteristics is well known. What is less appreciated is that it also possesses exemplary thermomechanical properties. The excellent mechanical properties of silicon are why silicon-based micro-machine technology has developed rapidly in recent years. These machines such as illustrated in FIG. 1 as optoelectric devices 20 are integrated devices called micro-electro-mechanical systems (MEMS). They may be composed of sensors, actuators, and/or electronic components all made out of silicon wafers using microfabrication techniques developed for the IC (integrated circuits) industry. Depending on their functions, these MEMS may range in size from a few microns to several hundred microns. Examples include micropumps, microvalves, and micro-acceleration sensors and related devices in a wide range of transportation, bio-medical, industrial process monitoring and control applications. The integration of TPS into many of these devices provides an on-board stand-alone micropower supply and also serves as the interconnect between the various electronic and sensor components. This combination can turn many current MEM's into self-powered smart (i.e. with problem-solving capability) micromachines, some of which may be optically transparent if the TPS is thin enough such as the few microns referred to above. It is known that the interior pore surface of porous silicon (PS) is extensively covered with hydride species. These include SiH, SiH.sub.2 and SiH.sub.3. It has been suggested that the dihydride may be the dominant one with a coverage on the order of 2.3.times.10.sup.14 cm.sup.-2. Also, PS is a material that has a specific surface area as high as 900 m.sup.2 cm.sup.-3. One can estimate the typical tritium content in TPS as follows. Assuming that the TPS synthesis processes converts 10% of the surface SiH.sup.2 into SiHT (chemisorbed) species, the concentration of SiHT (and hence of tritium) will be of the order of 2.1.times.10.sup.20 /c.c. This is equivalent to about 10 Ci cm.sup.-3. For the present example a TPS film of area 1 cm.sup.2 and 10 microns in thickness is considered. This film contains 10 mCi of tritium; however, TPS having radiation in the range of from about 10 mCi to 100 mCi is preferred. Conventional gas tube-type tritium-based lighting technology has a typical conversion capacity of around 0.4 ft.-L per Ci. As stated, TPS should have significantly higher conversion capacity than in this example. This is due mainly to the much higher tritium concentration that can be achieved in TPS compared to the gas phase, as well as the total accessibility of the luminescent centers, to the emitted beta particles from the tritium. As a result, the conversion efficiency for TPS is estimated to be of the order of 34 ft.-L per Ci, substantially above that of the conventional technology. Thus, this thin film of TPS material (with a tritium content of about 10 mCi per c.c.) can generate initially 0.34 ft-L of light. For comparison an average TV monitor emits at a level of about 2 ft-L. For low light applications such as emergency signs and instrumentation panel lighting, the brightness requirement can be considerably below that of TV monitor conditions. Accordingly, TPS is useful for stand-alone low level lighting applications either directly as a light source or as an optical power source for optoelectronic devices. In traditional tritium lighting devices the tritium is in gaseous form contained in a gas tube whose surface is coated with phosphor-type materials. Due to the short mean free path of the emitted beta particles one cannot increase arbitrarily the output light intensity by increasing the tritium partial pressure and/or the diameter of the tube. This kind of approach is self-limiting by the self-absorption effect from the short mean free path. The typical tritium content utilized in gas-tube type of illumination devices is about half a curie, generating light at a level of around 0.2 ft.-L, which give a conversion efficiency of around 0.4 ft.-L. per Ci. Now the tritium concentration achievable in TPS can be easily an order of magnitude or more above that of the gas phase. Above it was indicated assuming a 10% coverage of the silicon-related sites during the formation of TPS would lead to a tritium concentration of about 10 Ci per cc. For comparison 0.5 atmosphere of tritium (which is the typical tritium content in conventional gas-tube devices) is equivalent to about 1.2 Ci per cc. This means that given the same volume of luminescent material for the tritium can act on, TPS would have a conversion efficiency of 0.4 times (10/1.18) ft.-L. per Ci. This arises from the higher tritium density in TPS, an expected result for a solid-state device. However, this is only one advantage (but not the only one) of TPS. In a gas-tube type device only a thin layer of the luminescent material is accessible to the beta particles from the tritium. That layer has a thickness of the order of 1 micron or less, i.e., the range of the beta particles in the solid state. In TPS the tritium atoms are all dispersed microscopically through the volume of porous silicon. All of the luminescent material (i,e., the silicon) is accessible to some tritium. This leads to another increase of the TPS conversion efficiency by a factor on the order of the ratio of the thickness of the TPS film to the range of the beta particles. For a TPS film thickness of 10 microns, the enhanced conversion efficiency of TPS would be a value on the order of 34 ft.-L. per Ci. This is substantially above what can be achieved with conventional devices. The fundamental reason is that in TPS one manages to pack more tritium into a greater amount of accessible luminescent materials than is possible with conventional gas-tube type technologies. In summary, the conversion efficiency of TPS has been approximately estimated with the following relations: EQU E.sub.TPS =E.sub.GAS (N.sub.TPS /N.sub.GAS) d.sub.TPS /1.sup.SOLID) Here E.sub.GAS is the conversion efficiency of gas-tube type tritium devices, N.sub.TPS is the tritium density in TPS, PA1 N.sub.GAS is the tritium density in the gas phase, PA1 d.sub.TPS is the thickness of the TPS film, PA1 1.sub.SOLID is the range of the beta particles in solid state materials. It must be stressed that the above equation can only give an order of magnitude estimate. In particular, it was assumed that the internal quantum efficiencies of porous silicon and conventional phosphor-type luminescent materials are comparable. This need not be considered as overly optimistic since many results on conventional porous silicon have indicated a very high internal quantum efficiency. Nevertheless, a precise determination of the optical performance of TPS is not yet available. Turning to the safety factors, tritium is a weakly emitting radionuclide for which the decay process is effected 100% through beta emission. To address the safety issues there are two important estimates to consider. One is the range of the beta particles in typical/window packaging materials that are likely candidates to be utilized to isolate the TPS from the external environment. The other issue is the dose on an individual due to failure of the window/packaging materials and the release of all the tritium into the air from the TPS-containing device. The range of a beta particle in a solid material is inversely proportional to its density. For most solid materials simple estimates give ranges that are under 1 micron. This means that a few microns thick window/packaging barrier is more than sufficient to stop the beta particles. Note that this is in addition to the presence of the porous silicon (PS) whose very purpose and configuration have already been designed to absorb as much of the beta emission as possible. The above considerations apply to normal operation. Considering a situation in which for whatever reason there is a complete breakage of the window/packaging material barrier and all the tritium embedded in the pore surfaces of the TPS material is instantaneously released into the air (a truly incredible event). The instantaneous total release of the tritium (using 10 mCi tritium content) is a worst case scenario. In reality the tritium atoms are bonded to the pore surfaces in the form of surface chemisorbed species such as hydrides (e.g., SiHT) or as hydroxide. The actual release mechanism is likely to be through a combination of the slower processes of thermal desorption and exchange. However, the instantaneous total release assumption in the following estimate is used to provide a conservative boundary. The form of the released tritium is assumed to be the oxide form which represents the worst biohazard case. The likely scenario for this to happen is total oxidation of the tritium content in the TPS (e.g., as in a fire). Assume that the tritium (10 mCi in all) has been released into a room of volume V m.sup.3 with no venting to the exterior. The maximum dosage on an individual is of the order 0.9 rem. This assumes that this person would stay in the unvented room for a certain minimum period after the release. For a room of size 20 m.sup.3 with no ventilation, this minimum period amounts to about 16 hours. For reference the International Commission on Radiological Protection (ICRP) maximum permissible annual dosage is 5 rem. In fact, implementation of straightforward but effective safety procedures such as adequate ventilation and tritium detector/alarm systems would significantly further reduce the potential maximum dosage that can be delivered. While there has been disclosed what is considered to be the preferred embodiment of the present invention, it is understood that various changes in the details may be made without departing from the spirit, or sacrificing any of the advantages of the present invention.
	summary
062597566	summary	BACKGROUND OF THE INVENTION The present invention relates to a method of operating a boiling water nuclear reactor and particularly relates to a control blade sequence pattern to optimize the BWR power control and hence obtain greater power and/or improved uranium fuel efficiency and economy. Control blades typically containing boron carbide provide reactivity control, i.e., control the total reactive power and power distribution. The control blades are generally in cruciform cross-section and are inserted between four associated fuel bundles or assemblies containing arrays of fuel rods. The blades enter the bottom of the reactor core and are movable into positions within the reactor core for reactivity control and positions withdrawn from the core. In modern BWRs, control blades are divided into two groups, generally known as "A sequence" and "B sequence." The control blades of the A sequence and the B sequence are referred to herein respectively as A or B control blades or A or B sequence blades. These two groups generally form a checkerboard pattern throughout and in plan view of the reactor core. An A control blade is conventionally at the center of the core. Thus, every other control blade in a first direction, i.e., an X direction, from the center of the core is an A sequence blade and every other blade in a Y direction, i.e., a second direction normal to the first direction, from the center of the core is an A sequence blade. Alternate blades in the X and Y directions between the A sequence blades are B sequence blades. With this typical arrangement, the A sequence blades are symmetric relative to the center of the core, while the B sequence blades are asymmetric. Control blades are also designated as deep or shallow inserted blades. Blades whose tips are inserted more than two-thirds into the core are referred to as deep blades, while those inserted less than one-third into the core are referred to as shallow blades. Deep blades are used to control total reactive power as well as the global radial power shape. Shallow blades are used to control the reactor axial power shape. Generally, blades are not inserted into the middle third of the core because they tend to create axial power distribution problems. Control blades are, of course, movable to the deep and shallow positions from positions totally withdrawn from the core. In conventional BWR operations, four basic control blade patterns, the control blades of which are designated at A1, A2, B1 and B2, are used to develop operating control blade pattern sequences throughout each cycle. The typical practice is to alternate use of two blade sequences during operation of the reactor. Each sequence of blades, for example, the A sequence, is used exclusively of the other sequence, e.g., the B sequence. Thus, only A sequence or B sequence blades, but not both, are inserted into the core at any given time. The non-inserted blades are disposed in withdrawn positions. The blades are moved in and out relative to the core on a scheduled basis to avoid power distribution difficulties within the fuel bundles. When employing this operational practice, the asymmetry associated with B sequence blades, however, causes the power distribution within the reactor to become asymmetric and more highly peaked to various locations, i.e., the radial power shape of the core tends to be skewed toward one corner or another corner, i.e., non-symmetric. This asymmetric peaking must be accommodated and reduced by design of the reactor fuel loading, necessarily reducing fuel efficiency. Problems associated with the B sequence blade asymmetry, however, have been partly alleviated through the use of control cell core (CCC) loading patterns. With the CCC design, low reactivity fuel assemblies are placed in the control cells so that control rod motion occurs adjacent only relatively low-power fuel. That is, CCC loading schemes place relatively low power fuel bundles around A sequence blades, only some of which are used for power control. Fresh fuel bundles which have higher power must not be loaded next to these blades because the blades must be inserted for longer periods. With the CCC design, those certain A sequence control blades located in the low-power control cells are utilized for reactivity and power distribution control purposes throughout the entire cycle. CCC loading, however, has limitations and drawbacks. As reactors are operated for long periods of time between refueling cycles, the number of fresh bundles inserted each cycle increases. When the fresh batch fraction is larger than 40%, the CCC loading becomes difficult and inefficient. Moreover, longer control periods associated with the use of CCC designs may contribute to a type of fuel failure associated with power peaking with the bundles. Accordingly, there is a need for a control blade sequencing pattern which avoids the asymmetry problem associated with the B sequence blades but does not require that the reactor fuel loading be limited as in the CCC design. BRIEF SUMMARY OF THE INVENTION In accordance with a preferred embodiment of the present invention, the blades of each of the A and B blade sequences, i.e., main blade groups A and B, are respectively divided into blade sub-groups, referred to as A1 and A2 and B1 and B2 blades. With a control blade A2 at the center of the core, every other control blade location in the core in the X and Y directions is an A2 blade. The first and second sub-groups of each main group are flanked by first and second sub-groups, respectively, of the other main group in the X direction. For example, the first and second sub-groups A1 and A2 of main group A are flanked by first and second sub-groups B1 and B2, respectively, of the main group B in the X direction, while the first and second sub-groups B1 and B2 of main group B are flanked by first and second sub-groups A1 and A2, respectively, in the X direction. The first and second sub-groups of each main group are flanked by second and first sub-groups, respectively, of the other main group in the Y direction. Thus, first and second sub-groups A1 and A2 of main group B are flanked by second and first sub-groups B2 and B1, respectively, in the Y direction, while the first and second sub-groups B1 and B2 of main group B are flanked by second and first sub-groups A2 and A1, respectively, in the Y direction. Thus, A1 and A2 control blades are symmetric relative to the center point of the core, whereas the B1 and B2 blades are asymmetric. In accordance with a preferred embodiment of the present invention and contrary to prior practice, selected A and B control blades are inserted into the core at the same time but with the B sequence blades used only as shallow blades and the A sequence blades used as either deep or shallow blades. That is, the B sequence blades are movable only between positions withdrawn from the core and shallow positions within the core. The A sequence blades, however, are movable between withdrawn, deep or shallow positions. Further, the operating sequence is such to allow each fuel bundle a period of operation with its associated control blade withdrawn that is at least twice as long as the previous period of operation with the associated control blade inserted. The pattern repeats itself every three or more consecutive time periods and any one blade is fully withdrawn from the core for at least two consecutive time periods after it has been inserted enabling operation of the associated fuel assembly in an uncontrolled state for those two consecutive periods. That is, a different one of control blades A1 or A2 are inserted from a withdrawn position into a deep position at the beginning of each of at least three successive time periods, while a different one of the control blades A1, A2, B1 or B2 is inserted into a shallow position at the beginning of each such successive time period. Thus, selected non-inserted blades are maintained in withdrawn positions at the beginning of each such successive time period for at least two consecutive time periods. In a preferred embodiment according to the present invention, there is provided in a nuclear reactor having a core, the core including plural control blades and fuel assemblies with each control blade operationally associated with four fuel bundles of each fuel assembly, sets of the control blades forming discrete control blade groups forming a core pattern including first and second main groups thereof, respectively, symmetrically and asymmetrically arranged about the core, each main group having first and second sub-groups, the first sub-groups of the main group including first and second operational sub-groups, respectively, of control blades, each first sub-group of the first main group being flanked by first sub-groups of the second main group in a first direction and by the second sub-group of the second main group in a second direction normal to the first direction, each second sub-group of the first main group being flanked by second sub-groups of the second main group in the first direction and by first sub-groups of the second main group in the second direction, each first sub-group of the second main group being flanked by a first sub-group of the first main group in the first direction and by the second sub-group of the first main group in the second direction, each second sub-group of the second main group being flanked by a second sub-group of the first main group in the first direction and by the first sub-group of the first main group in the second direction, each blade of each first main group being movable between withdrawn, deep and shallow positions relative to the core and the fuel bundles associated with the first main group, each blade of each second main group being movable between withdrawn and shallow positions relative to the core and to fuel bundles associated with the second main group, a method of operating the nuclear reactor comprising the steps of inserting into the core from a withdrawn position into a deep position a different sub-group of the blades of the first and second sub-groups and operational sub-groups of the first main group at the beginning of at least each of three successive time periods, respectively, with predetermined sub-groups of blades of the first and second main groups withdrawn from the core at the beginning of each successive time period, inserting into the core from a withdrawn position into a shallow position a different sub-group of the blades of the first and second sub-groups of the first and second main groups at the beginning of each of the successive time periods with the predetermined sub-groups of blades of the irst and second main groups withdrawn from the core at the beginning if each successive time period and maintaining selected sub-groups of the withdrawn blades at the beginning of each successive time period in a withdrawn position for at least two consecutive time periods. In a further preferred embodiment according to the present invention, there is provided in a nuclear reactor having a core, the core including plural control blades and fuel assemblies with each control blade operationally associated with four fuel bundles of each fuel assembly, sets of the control blades forming discrete control blade groups forming a core pattern including first and second main groups thereof, respectively, symmetrically and asymmetrically arranged about the core, each main group having first and second sub-groups, each sub-group at each control blade location within the core being located alternately in both orthogonally-related X and Y directions in plan view of the core, the first sub-group of the first main group including an operational sub-group, a method of operating the nuclear reactor comprising the steps of inserting into the core from a withdrawn position into a deep position different selected blades of the first and second sub-groups of the first main group at the beginning of at least each of three successive time periods, respectively, with predetermined sub-groups of blades of the first and second main groups withdrawn from the core at the beginning of each successive time period, inserting into the core from a withdrawn position into a shallow position different selected blades of the first and second sub-groups of the first and second main groups at the beginning of each of the successive time periods with the predetermined sub-groups of blades of the first and second main groups withdrawn from the core at the beginning if each successive time period and maintaining selected withdrawn blades at the beginning of each successive time period in a withdrawn position for at least two consecutive time periods.
	description	This invention relates to a device and process for switching and controlling an electron dose emitted by a micro-emitter, for example a microtip. Microtip type micro-emitters will be considered in the remainder of the description, as a non-restrictive example. The subject of microtips, now accompanied by the subject of nanotubes, defines a range of applications for FED (Field Emission Display) displays and also for micro-emitters, in which requirements in terms of switching and controlling of emitted flows are very severe. In the case of a hot emission (diodes, triodes, cathode ray tubes), electrons acquire sufficient energy (called “output work”) due to their thermal agitation to go beyond the potential barrier that retains them to nuclei. They are then moved towards the material surface and, if there is an electric field that attracts them, they can be extracted from this material. At ordinary temperatures, the thermal agitation energy is not sufficient for electrons to exit from the material. In the case of a cold emission, based on the principle of a field effect test in a vacuum chamber, a tunnel effect enables electrons to be extracted from the emitter (cathode) in the vacuum and then to be collected on an anode. Emitters working in cold emission are considered as being voltage controlled current sources, the flow of emitted electrons obeying Fowler-Nordheim equations. For example, this is the case of a microtip 10 made of Tungsten used as an electron emitter. Its electrical scheme is shown in FIG. 1A. An electron flow is set up between the anode 11 and the cathode 12. A control voltage is applied between the extraction grid 13 called the “gate”, and the cathode 12. FIG. 1B shows the behavioral symbol of such a microtip 10 that can be used with a generic electrical simulator (“Spice” type) The emission condition for such a microtip 10 is characterized by strong non-linearity of the emission current Itip as a function of the voltage applied on the extraction grid 13. The tip current Itip satisfies the law:Itip=afnV2exp−bfn/Vgc The coefficients afn and bfn depend on the geometric characteristics of the microtip. One such current-voltage characteristic is illustrated in FIG. 2. An example of an operating point (Itip=Ion for Vgate−cathode=Von) is shown in this figure. The ideal characteristic is shown as reference 14. In reality, this type of characteristic cannot be reproduced from one microtip to another. The result is curves 15 shown in dashed lines. Therefore one of the disadvantages of cold emission is to reveal some instability in the value of the current, which is equivalent to a noise generated by output working fluctuations inherent to local surface contaminations. These fluctuations are variable from one microtip to another and are also variable in time for the same microtip. There are two possible types of microtip control: a current control by a current regulation device: this type of possibility is used in FEDs (Field Emission Display) through a single or “multigate” transistor located in series in the cathode circuit, as described in document references [1] and [2] at the end of the description. The current emitted by each microtip may be programmed theoretically. It is independent of the quality and characteristics of each microtip. The voltage Vgc is modulated from one microtip to another or in time. One of the defects of such a control is that it mixes a low voltage (LV) and a high voltage (HV) at the transistor switching and controlling circuit, because the extraction electrode must be increased to a few tens of Volts. The visual display matches the limited operating precision frequency of this type of control. a voltage control: if care is not taken, the emission current will be modulated which may be unacceptable for some applications. If the current excursion and particularly the extreme values are known, and if the quantity to be controlled is the electric charge, this type of solution is satisfactory when it is combined with a variable observation time window, Tnom. Q = I nom * T nom = 2 ⁢ I nom * T nom 2 = I nom 2 * 2 ⁢ T nom The device according to the invention is a circuit of this type that is naturally faster and in which the observed linearity defects are corrected, the extraction grid HV control circuits being independent of the LV circuits controlling the electric charge, which simplifies use of the circuit and reduces the sensitivity to noise. Therefore several solutions are possible to measure the quantity of electrons transmitted by a microtip. In some cases, it is possible to work in current regulation as illustrated in FIGS. 3A and 3B. Emission of a calibrated current (generator 16) for a given time delimits an electric charge according to the law Q=T.t. This type of current regulation system includes a sensitive tip current detection element 17, a reference current test element 18 and a current adjustment element 19. This system may operate: In open loop in the case of a sequential calibration, then programming of a given number of measurements with a single reference, as illustrated in FIG. 3A, In closed loop in the case of a current servocontrol in real time as illustrated in FIG. 3B, and as described in document reference [3]. In the embodiment illustrated in FIG. 3A, the specification for the system must allow the necessary time to perform the calibrations. This type of implementation is incapable of correcting imperfections in the electron beam for which the recurrence frequency is higher than the calibration refreshment frequency. In the embodiment shown in FIG. 3B, the stability of the counter-reaction loop is essential and it must be guaranteed, usually at the price of active compensation of the pass band of the looped system and therefore to the detriment of its speed performances. Requirements in terms of speed, stability, noise and linearity make it impossible to use this type of implementation in many applications. A global method of controlling weak electric charges consists of defining the required quantity of charges, interrupting the electron beam when the required dose has been reached (“dose control”), using several configuration input variables. In this case, the quantity of electric charges is defined in advance. The device used for this control must operate on a tip current dynamic, particularly including current fluctuations in time for the same microtip. Theoretically, this type of method enables very good linearity. However, the use of real functional modules and the requirement for high frequency operation result in strong non-linearities in the electric charge controlled as a function of the current state. A document reference [4] according to known art mentioned at the end of the description describes a two-dimensional network of miniature cathodes used as electron beam emitters that are numerically addressable. This network includes internal electron focusing for each emitter, a closed loop electron dose control circuit for controlling each emitter by precisely controlling the electron flow. This type of dose control circuit connected to an emitter can be used to obtain a dose delivered during each cycle write, adapted despite emitter-to-emitter mismatch, temperature and ageing effects. This control circuit terminates the emission at a fixed dose rather than at a fixed time. It is an integrated component and is connected to the emitter. But this type of control circuit is a source of non-linearities. Furthermore, for a linear or two-dimensional arrangement of microtips, it cannot compensate for dispersions of doses emitted due to current dispersions inherent to microtips. The purpose of the invention is to compensate for this type of non-linearity so as to make the control device linear and useable, and to provide specific solutions for linear or two-dimensional devices. The invention relates to a switching and controlling device for an electron dose emitted by a micro-emitter, for example a microtip, characterized in that it comprises: a sensor module that receives the output current from the micro-emitter and a voltage to adjust the polarization point of the said device, a comparator module that receives the output signal from the said sensor module, and a threshold voltage to adjust the quantity of electrons to be emitted, a logical module that receives the output signal from the comparator module, and a start signal to initialize the electron emission, and a logical signal to define whether or not the micro-emitter should emit, a control module that receives the output signal from the said logical module that generates the voltages necessary for initialization and extinction of the micro-emitter current pulse, means of varying the threshold voltage such that the sum S=Nstart+Nmeasure+Noff remains constant during the electron emission, where Nstart is the number of electrons at the current pulse start time, Nmeasure is the number of electrons at the measurement time of this current pulse, Noff is the number of electrons at the extinguishing time of this current pulse. In a first example embodiment, the device according to the invention comprises means of modulating the threshold voltage in time starting from the initialization signal so as to program a variable dose control in time such that excess electrons emitted during the initialization and extinguishing times are strictly compensated by a reduction of the programmed dose in time. In a second embodiment, the device according to the invention also comprises: a module for detecting the micro-emitter current, capable of reproducing the tip current Itip exactly, or adding a gain on the current, a variable voltage generation module that outputs a set voltage V2=f(Itip). The invention also relates to a linear or matrix switching and controlling device for electron doses emitted by a set of micro-emitters, characterized in that it comprises the following for each micro-emitter: a sensor module that receives the output current from the micro-emitter and a voltage to adjust the polarization point, a comparator module that receives the output signal from the said sensor module and a threshold voltage to adjust the quantity of electrons to be emitted, a logical module that receives the output signal from the comparator module, and a start signal to initialize the electron emission, and a logical signal to define whether or not the micro-emitter should emit, a control module that receives the output signal from the said logical module that generates the voltages necessary for initialization and extinction of the micro-emitter current pulse, means of varying the threshold voltage such that during the electron emission, the sum S=Nstart+Nmeasure+Noff remains approximately constant, where Nstart is the number of electrons at the current pulse start time, Nmeasure is the number of electrons at the measurement time of this current pulse, Noff is the number of electrons at the extinguishing time of this current pulse. The invention also relates to a process for switching and controlling an electron dose emitted by a micro-emitter comprising: a step to convert the current output by the micro-emitter and to adjust the operating polarization point, a step to compare the signal obtained at the output from the previous step with a threshold voltage for adjustment of the electron quantity to be emitted, a logical step to initialize the electron emission, and to define whether or not the micro-emitter should emit, a control step that generates the voltages necessary for initialization and for extinction of the micro-emitter current pulse, characterized in that it comprises a step to vary the threshold voltage such that during the emission of electrons, the sum S=Nstart+Nmeasure+Noff remains constant, Nstart being the number of electrons at the current pulse start time, Nmeasure being the number of electrons at the measurement time of this current pulse, Noff being the number of electrons at the extinguishing time of this current pulse. This type of invention has a wide field of applications: electron emission by cold cathode, switching and controlling of weak electric charges, compensation of charge measurement errors, high operating frequency, solution compatible with application specific integrated circuits (ASICs). The switching and controlling device of an electron dose emitted by a micro-emitter illustrated in FIG. 4 is composed of a microtip 10 with an anode 11, a cathode 12 and an extraction grid 13, capable of supplying a current when the voltage of the extraction grid 13 relative to the cathode 12 becomes greater than the extraction voltage in the vacuum. Parasite capacitances 20 and 21 are inherent to the fabrication of such a microtip 10 in microtechnology. This device comprises: a sensor module 30 that performs an electron-voltage conversion and that receives the current Ic output by this microtip 10 and a voltage V1 to adjust the polarization point of the said device, the sensitivity of a module being expressed in Volts/electrons, a comparator module 31 that receives the output signal Vse from the said sensor module 30 and a threshold voltage V2 to adjust the quantity of electrons to be emitted, and that outputs a sufficient charge detection signal Vcom, a logical module 32 that receives this signal Vcom, and a Start signal to initialize the electron emission, and a logical data signal to define whether or not the microtip should emit, a control module 33 that receives the output signal from the said logical module 32 and the Vg-on and Vg-off signals that generate the voltages necessary for initialization and extinction of the microtip current pulse (several tens of Volts). This device is actually applicable to an arrangement of several microtips either in the form of a linear arrangement (strip) or a two-dimensional arrangement (matrix). All combinations of arrangements are also possible. This device can be made using a specific high voltage technology, and can control electron doses emitted at high rates. We will now analyze each of these modules 30, 31, 32 and 33. The role of this module 30 is to process the basic information available on the microtip 10 and to convert it into a magnitude that can be compared with an input magnitude, in order to take a decision on the number N of electrons emitted. This module may advantageously be composed of a CTIA (Capacitive TransImpedance Amplifier) amplifier that makes a current—voltage conversion. The input variable is then the cathode current of the microtip Ic. This amplifier is characterized by its conversion gain that is expressed in Volts/e−. It is composed of an amplifier 35, a counter reaction capacitor (Cfb) 36 and a reset device 37. The result for the output excursion ΔVs of the sensor module is: Δ ⁢ ⁢ V s = - I c * T i ⁢ ⁢ n ⁢ ⁢ t C f ⁢ ⁢ b = N * q e C f ⁢ ⁢ b = N ⁢ ⁢ ℛ This type of solution is advantageous compared with a solution making a direct integration on the microtip capacitance for several reasons: the signal is not sensitive to parasite capacitances on the input side, its conversion gain may be fixed precisely. It is defined by the value of Cfb. For example, it may be 23 μV/e− for Cfb=7 fF, the cathode polarization point is fixed by the external variable V1. This module 31 receives two analog voltages on its inputs: the output voltage Vsc from the sensor module 30, the control voltage V2 that fixes the value of the comparison threshold. This module comprises an amplifier 40 in open loop, for which the output level comprises two states (VDD and VSS) equivalent to two logical states as a function of the input voltages: as long as Vsc>V2, the logical output Vcom remains equal to “1”, when Vsc=V2, the logical output Vcom switches and is set to a logical “0”. This module 32 has several internal signal sequencing and generation functions. Its roles are to: latch the decision made Vcom obtained at the output from the comparator module 31 until the arrival of a reset signal, generate non-overlapping phases useful for resetting the sensor module 30 and the control module 33. This module is initialized by a start signal at the beginning of the sequence, and obeys the data signal as illustrated in the following table: DataAction1Emission from the microtip0No emission from the microtip This module 33 establishes the extraction grid voltage necessary for the microtip to emit the required current synchronously with the appearance of the start signal. When the emitted electron dose has been reached (decision signal Vcom emitted by the comparator module 31), this module 33 cuts off the flow by bringing the extraction grid voltage to a level such that the electron current is reduced by several decades. These ignition and extinguishing values depend on the transconductance of the microtip and its geometric model. Control voltages may be switched from 20 V to about 50 V, which then requires the use of a specific high voltage technology (HVCMOS). The main function of this module 33 is therefore to translate the level [0-3 V] to [20 V-50 V]. This type of switching and controlling device has many limitations, inherent to the principle used. The voltage Vse obtained at the output from the sensor module 30 is proportional to the cathode current Ic emitted by microtip. Considering V1 as being the initialization voltage level, the number Ne of electrons emitted by the microtip is such that: N e = Q e q = ( V s ⁢ ⁢ e - V1 ) ℛ ⁢ ⁢ where ⁢ ⁢ ℛ = q C f ⁢ ⁢ b Qe is the electric charge emitted and q is the charge of the electron. Therefore, a calibrated charge Qc may be programmed by V2 with the following relation: N ⁢ ⁢ c = Q c q = ( V2 - V1 ) ℛ The value of the comparison threshold V2 fixes the programmed electric charge. If all modules were perfect, the sensor module 30 would immediately transmit a representation Vse of the cathode current Ic, the comparator module 31 would not have any delay, and the extraction grid control would instantaneously activate making or breaking the electron flow, according to the time diagram in FIG. 7A. Regardless of the level of the electron current, the emitted charge would be identical, and as illustrated in FIG. 7B: a nominal current ICnom would be interrupted after a certain time Tnom, a current 2ICnom would be interrupted after a time tnom/2, a nominal current 0.5Icnom would be interrupted after a time 2tnom. The areas shown in grey in each of the three cases are equal. In reality, the global duration of the current pulse is not linear as a function of the programmed current level. Due to the parasite capacitances 20 and 21 mentioned above, switching of the extraction grid 13 by several tens of Volts temporarily disturbs the input of the sensor module 30 for which the polarization has to be maintained to prevent saturation. This type of saturation would then require a large time constant before a return to equilibrium and would not enable operation at high frequency. During this time in which polarization of the sensor module 30 is maintained to creation of the electron flow, electron charges are already emitted and need to be counted in the global balance of emitted charges, although they cannot be measured since they depend on the current level that is not known in advance. This type of phenomenon is a first source of non-linearities. Another phenomenon occurs when the electron beam is extinguished, when Vse reaches V2. The comparator module 31 has a delay in making the decision, which is inherent to any electron module. During this delay, the microtip 10 continues to emit and therefore there is an additional extinguishing charge that is added into the global balance of the emitted charges. FIG. 8 shows a materialization of the error on the number N of programmed electrons, and illustrates such a phenomenon. If the number of electrons emitted is plotted as a function of time with respect to the number of programmed electrons, with a constant delay, an error is observed on the number of electrons emitted depending on the current level. In this figure, curve 45 corresponds to 2Iinom, curve 46 corresponds to Iinom and curve 47 corresponds to Iinom/2, curve 48 corresponds to the number of electrons emitted. Therefore, there is an overshoot on the charge emitted with respect to the programmed charge, which is a second source of non-linearities. A first solution for compensating for such non-linearities uses a comparison threshold that varies as a function of time. To achieve this, all that is necessary is to send a stair-case 50 on the input V2 of the comparator module 31 as illustrated in FIG. 9. The purpose of the invention is to compensate for such non-linearities by proposing other compensation methods by controlling the cathode current in Ic and by feedback on the value of the threshold V2. By analyzing the profile 55 of the microtip current pulse, it can be decomposed into a series of elementary times t1 to t6: t1: time to set up the voltage Vgate + reset CTIA, t2: latching time for the CTIA reset to cancel charge injection effects and transients, t3: measurement time, t4: comparator decision making delay time, t5: delay time due to cutoff of Vgate (logical), t6: delay to stop the electron flow. Some of these elementary times can be grouped together, to give the following simplified model: t1+t2=tstart: initialization time that extends from tdébut (corresponding to the beginning of the pulse) until tstart-control (corresponding to the effective beginning of the dose control), t3=tmeasure: actually controllable measurement time that extends from tstart-control until tstart-control (corresponding to the end of the dose control), t4+t5+t6=toff: extinguishing time that extends from tend-control until tfin corresponding to the effective end of the dose emission. If it is considered that the current reaches its nominal value Isteady-state quickly during the initialization time tstart and that it is latched for an extinguishing time toff, as a first approximation it is therefore constant during the entire duration of the current pulse. At the beginning, the setup time for Vgate is short and at the end, logical and extinguishing delays for Vgate are largely dominated by the delay of the comparator module 31 when the decision is being made. The total dose emitted as a number of electrons can be expressed as follows: N beam = N measure + I steady ⁢ - ⁢ state * ( t start + t off ) q e where N measure = ( V 2 - V 1 ) ℛ ⁢ ⁢ where ⁢ ⁢ ℛ = q e / Ctia The predicted electron dose is fixed by Nmeasure, but in fact an excess dose can be added due to non-zero start and extinguishing times. FIG. 12 illustrates a curve showing the number of electrons emitted as a function of the current state. In theory, as mentioned above, the number of electrons emitted should remain the same regardless of the current Itip, as illustrated by the horizontal curve 56. Curves 57 and 58 illustrate the number of electrons emitted during the initialization and the extinguishing times respectively. The sequencing may be such that the times tstart and toff remain constant regardless of the current, in other words the electrons emitted during these times tstart and toff only depend on the current state (affine function). The number of emitted electrons appears on the curve 59 which, for any value of the abscissa X, represents the sum of the curves 56+57+58. The relative numeric indication obtained from these curves shows an error on the number of electrons emitted with respect to the set value by a factor of 1.3 to 2.6. This is not acceptable for the required emission control precision. The purpose of the device according to the invention is to be capable of precisely emitting a programmed number of electrons regardless of the current state of the microtip and to interrupt the electron beam as soon as this value has been reached. Therefore the sum of electrons emitted during each of the times described above must remain constant, i.e., the total number of electrons emitted must be linear and constant regardless of the tip current Itip. The law for variation of the number of electrons emitted during the initialization and extinguishing times for the current pulse (affine function) is known. Therefore, it is possible to act on the test of the number Nmeasure of electrons effectively measured such that the sum S=Nstart+Nmeasure+Noff remains constant. In fact, Nmeasure therefore needs to decrease when Itip increases. To achieve this, the value of the threshold detection voltage V2 is modified during the electron exposure. Compensation is made on excessive electron quantities satisfying the following law: I tip * t q e Two types of compensation are possible: a time compensation or a compensation as a function of the current. FIGS. 13A and 13B illustrate theoretical curve 60 and measured curve 61 respectively, and theoretical curve 60 and measured curves 61′ of the relative number of electrons as a function of the tip current Itip respectively, without compensation and with compensation respectively, as a function of the current. Curve 61′ demonstrates the improvement to be obtained by using such an active compensation as a function of the current. FIG. 13B shows the stability of the number of electrons emitted as a function of the tip current, although there is an offset that remains inherent to the method used. The time denoted tmeasure cannot be zero since in this case nothing would be tested. The minimum time necessary for the compensation to work correctly must be such that the noise added by the sensor module 30 remains weak compared with the signal being processed by this module (typically Noffset=400 electrons, or ΔVS—min=8 mv). The invention also relates to a linear or matrix switching and controlling device for electron doses emitted by a set of micro-emitters, that comprises different modules 30, 31, 32 and 33 and means of varying the threshold voltage as described above, for each micro-emitter. Time Compensation This type of compensation is illustrated in FIG. 14. It does not cover all needs. It is capable of compensating for disparities between microtips, but not high frequency fluctuations on the same microtip. However, it can be used as soon as it is certain that the recurrence frequency of current fluctuations is less than the frequency of appearance of the programmed pulses. The threshold voltage V2 is modulated in time starting from the start signal so as to program a dose control variable in time such that excess electrons emitted during the tstart and toff phases are precisely compensated by the reduction of the programmed dose with time. Programmed ⁢ ⁢ dose = N prog =  ( V2 ⁡ ( t ) - V1 )  ℛ ⁢ ⁢ where ⁢ ⁢ ⁢ ℛ = q e C tia This time variation is controlled by the generator 65. Active Compensation as a Function of the Current When the frequency of current fluctuations is such that the current can vary for an elementary exposure time, the previous time correction is no longer sufficient. In the expression of the balance of the number of electrons emitted: N e - = I tip * T q e The two variables Itip and T vary simultaneously during the test. Therefore, it is no longer possible to test one of the variables while measuring the other. An active correction is necessary as a function of the current. FIG. 15 illustrates a simplified compensation diagram as a function of the tip current. A tip current detection module 67 is capable of precisely reproducing the tip current or introducing a gain (X) on this current, for example using a current mirror. This output current is measured by the sensor module 30. The input current Itip is also used as a reference for the variable voltage generation module 68 that outputs a set voltage V2=f(Itip). The decision on the time is always taken by the comparator module 31, but the decision threshold V2 is indexed on the instantaneous value of the emission current. The result is thus optimum compensation. More precisely, using the same notations as in FIG. 11, the number of electrons emitted in each of the phases can be calculated: Initialization phase N start = q ⁢ I * ( t début_contrôle - t début ) q Measurement phase N measure = q ⁢ ⁢ I * ( t fin_contr ⁢ o ^ ⁢ le - t d ⁢ e ' ⁢ but_contr ⁢ o ^ ⁢ le ) q Extinction phase N stop = q ⁢ ⁢ I * ( t fin - t fin_contr ⁢ o ^ ⁢ le ) q The number of electrons deposited in excess to be compensated by modifying the voltage V2 is equal to: N start + N stop = ⁢ I q [ ( t d ⁢ e ' ⁢ but_contr ⁢ o ^ ⁢ le - t d ⁢ e ' ⁢ but ) + ( t fin - t fin_contr ⁢ o ^ ⁢ le ) ] = ⁢ I q [ t start + t off ] Hence as a function of ΔV2: N start + N stop = C * Δ ⁢ ⁢ V2 q namely Δ ⁢ ⁢ V2 = I C [ t start + t off ] Since the capacitance of the sensor block and times └tstart+toff┘ are known, the variation of V2 to be programmed is directly proportional to I. The voltage difference to be programmed with respect to Vref (voltage to be applied to obtain the required dose during the measurement phase if Nstart and Nstop did not exist), can therefore be used, for example using a resistance RL to set up a voltage RLI, where RL=(tstart+toff)/C. In the special case in which the CTIA amplifier is recharged to a high state, this voltage RLI must be added to the voltage Vref to stop supply the microtip and therefore its emission, more quickly than in the ideal case (without Nstart and Nstop). For example, the block 68 in FIG. 15 can then be made as illustrated in FIG. 16. The transistor dimensions are chosen to satisfy the specified function in a manner known by those skilled in the art. This type of embodiment is advantageous in the sense that it enables carrying out all required functions close to or in the electron emission site, which has several advantages: it individually compensates for non-uniformities in emission of microtips or any other device, it performs these various functions in an ASIC (Application Specific Integrated Circuit), consequently, it participates in improving production efficiencies of microtips and their life, it is directly possible to access large two-dimensional emitters without making the various peripheral interfaces more complex (automatic processing of the in-pixel signal). [1] “Structure optimisation of transistor-based Si field emitter arrays” by T. Matsukawa, K. Koge, S. Kanemaru, H. Tanoue and J. Itoh (TIDW'98, pages 671-674, FED 2-4) [2] “Active matrix field-emitter arrays for the next-generation FEDs” by J. Itoh, S. Kanemaru, T. Matsukawa (199, SID) [3] U.S. Pat. No. 6,392,355 B1 [4] “Digital electrostatic electron-beam array lithography” by L. R. Baylor, D. H. Lowndes, M. L. Simpson, C. E. Thomas, M. A. Guillorn, V. I. Merkulov, J. H. Whealton, E. D. Ellis, D. K. Hensley, A. V. Melechko (J.Vac.Sci.Technol. B20 (6), November-December 2002, pages 2646-2650)
	summary
	claims	1. A multi leaf collimator (MLC) comprising:a plurality of leaves having a length of travel, wherein each leaf comprises:a blocking portion having a length L and a width W and being radio opaque, the blocking portion configured to block treatment level radiation,a drive portion having a length L′ and a width W′, the drive portion connected to the blocking portion, the drive portion including a recessed portion, anda conductive coil operatively connected to an electrical current source, whereinthe conductive coil is positioned in the recessed portion and fixed to the drive portion along at least a portion of length L′, and wherein electrical current passing through the conductive coil generates a first magnetic field;a leaf guide configured to support the plurality of leaves; anda plurality of stationary magnets, each magnet positioned adjacent to the drive portion of at least one leaf, wherein each stationary magnet has a second magnetic field configured to operate in conjunction with the first magnetic field to exert a force on the drive portion. 2. The MLC of claim 1, wherein the MLC comprises a first bank including the plurality of leaves and a second bank including the plurality of leaves. 3. The MLC of claim 1, further comprising a first leaf position encoder. 4. The MLC of claim 3, further comprising a second leaf position encoder. 5. The MLC of claim 1, wherein the width W of the blocking portion is greater than the width W of the drive portion. 6. The MLC of claim 5, wherein the length of each of the plurality of stationary magnets (L″) is selected such that the number of coils adjacent to the plurality of stationary magnets at any given time during leaf motion remains approximately constant. 7. A multi leaf collimator (MLC) comprising:a plurality of leaves having a length of travel, wherein each leaf comprises:a blocking portion having a length L and a width W and being radio opaque, the blocking portion configured to block treatment level radiation,a drive portion having a length L′ and a width W′, the drive portion connected to the blocking portion, anda conductive coil operatively connected to an electrical current source, wherein the conductive coil is fixed to the drive portion along at least a portion of length L′, and wherein electrical current passing through the conductive coil generates a first magnetic field;a leaf guide configured to support the plurality of leaves;a plurality of stationary magnets, each magnet positioned adjacent to the drive portion of at least one leaf, wherein each stationary magnet has a second magnetic field configured to operate in conjunction with the first magnetic field to exert a force on the drive portion;a magnetic drive module, the magnetic drive module including:an upper portion having the drive portion of a first subset of the plurality of leaves extending therein and a first subset of the plurality of stationary magnets, wherein each stationary magnet of the first subset is positioned on either side of the drive portion of the first subset of the plurality of leaves;a lower portion having the drive portion of a second subset of the plurality of leaves extending therein and a second subset of the plurality of stationary magnets, wherein each stationary magnet of the second subset is positioned on either side of the drive portion of the second subset of the plurality of leaves;wherein the first subset of stationary magnets in the upper portion are horizontally offset from the second subset of stationary magnets in the lower portion; andwherein the drive portion in the first subset and second subset of the plurality of leaves is alternatingly located in the upper portion and the lower portion for each adjacent leaf. 8. The MLC of claim 7, wherein the first subset of stationary magnets in the upper portion are horizontally offset from the second subset of stationary magnets in the lower portion by approximately a thickness of one leaf. 9. A system for collimating a therapeutic radiation beam, the system comprising:a multi leaf collimator (MLC) comprising:a plurality of leaves having a length of travel, wherein each leaf comprisesa blocking portion having a length L and a width W and being radio opaque, the blocking portion configured to block treatment level radiation,a drive portion having a length L′ and a width W′, the drive portion connected to the blocking portion, the drive portion including a recessed portion, anda conductive coil operatively connected to an electrical current source, wherein the coil is positioned in the recessed portion and fixed to the drive portion along at least a portion of the length L′, and wherein electrical current passing through the coil generates a first magnetic field;a leaf guide configured to support the plurality of leaves; anda plurality of stationary magnets, each magnet positioned adjacent to the drive portion of at least one leaf, wherein each stationary magnet has a second magnetic field configured to operate in conjunction with the first magnetic field to exert a force on the drive portion; anda driver component, wherein the driver component directs electrical current to the coil, thereby causing movement of the plurality of leaves to desired states. 10. A method for collimating a radiation beam with a multi leaf collimator (MLC), the method comprising:determining a desired state for a leaf of the MLC, wherein the leaf is configured to block treatment level radiation;activating a magnetic field of a conductive coil positioned in a recessed portion of the leaf, the leaf having a length of travel;if the leaf is not in the desired state, modifying the magnetic field of the conductive coil to interact with a second magnetic field and result in a force on the leaf causing the leaf to move; andstopping the leaf at the desired state anywhere along the length of travel. 11. The method of claim 10, wherein the leaf moves at a speed of at least 50 cm/s. 12. The method of claim 10 further comprising:wherein modifying the magnetic field of the conductive coil when the leaf is not in the desired state includes applying an electrical current to the conductive coil residing within a driving portion of the leaf to generate a first magnetic field, wherein the first magnetic field operates in conjunction with the second magnetic field of stationary magnets on either side of the driving portion resulting in the force on the driving portion causing the leaf to move. 13. The method of claim 12, wherein stopping the leaf at the desired state comprises applying a braking force by modifying the electrical current to the conductive coil such that the first magnetic field operates in conjunction with the second magnetic field resulting in a braking force to the driving portion of the leaf. 14. The method of claim 10, wherein the radiation beam comprises one of photons, protons, and other ions. 15. A multi leaf collimator (MLC) comprising:a plurality of leaves having a length of travel, wherein each leaf comprisesa blocking portion having a length L and a width W, wherein the blocking portion is radio opaque and configured to block treatment level radiation; anda drive portion having a length L′ and a width W′, the drive portion connected to the blocking portion, the drive portion including a recessed portion, anda conductive coil operatively connected to an electrical current source, the conductive coil positioned in the recessed portion and fixed to the drive portion along at least a portion of length L′; andwherein at least one of the leaves is capable of moving at a speed of at least 50 cm/s and stopping at a position anywhere along the length of travel. 16. The MLC of claim 15, wherein at least one of the leaves is capable of stopping anywhere along the length of travel to within 100 microns or less of a desired position. 17. A multi leaf collimator (MLC) comprising:a plurality of leaves having a length of travel, wherein each leaf comprisesa blocking portion having a length L and a width W, wherein the blocking portion is radio opaque and configured to block treatment level radiation; anda drive portion having a length L′ and a width W′, the drive portion connected to the blocking portion;wherein at least one of the leaves is capable of moving at a speed of at least 50 cm/s and stopping at a position anywhere along the length of travel;a plurality of conductive coils operatively connected to an electrical current source, wherein each of the conductive coils is fixed to the drive portion of one of the leaves along at least a portion of length L′, and wherein electrical current passing through the conductive coils generates a first magnetic field;a leaf guide configured to support the plurality of leaves; anda plurality of stationary magnets positioned adjacent to the drive portion, wherein each stationary magnet has a second magnetic field configured to operate in conjunction with the first magnetic field to exert a force on the drive portion. 18. The MLC of claim 17, wherein each of the plurality of stationary magnets is positioned one on either side of the drive portion. 19. A multi leaf collimator (MLC) comprising:a plurality of leaves having a length of travel, wherein each leaf comprises:a blocking portion having a length L and a width W and being radio opaque, the blocking portion configured to block treatment level radiation,a drive portion having a length L′ and a width W′, the drive portion including a recessed portion and connected to the blocking portion, anda permanent magnet positioned in the recessed portion of the drive portion;a leaf guide configured to support the plurality of leaves; anda plurality of conductive coils, at least one coil positioned between adjacent leaves and connected to an electrical current source to generate a first magnetic field when current passes through the at least one coil that interacts with a second magnetic field generated by the magnet to exert a force on the drive portion. 20. The MLC of claim 19, wherein the width W of the blocking portion is greater than the width W′ of the drive portion.
	summary
043691617	claims	1. Rack and pinion mechanism for moving a unit absorbing neutrons movable vertically in a nuclear reactor for controlling the power and for emergency shutdown of said reactor, comprising (a) a vertical control shaft whose lower part is connected to said absorbent unit and whose upper part constitutes a rack disposed in the longitudinal direction of said shaft; (b) a first pinion rotatably driven by drive means, in engagement with said rack for moving said control shaft and said absorbent unit in both directions; (c) said first pinion being in permanent engagement with said rack and being mounted to rotate on a shaft connected to said drive means for setting said first pinion in rotation perpendicularly to said control shaft, said first pinion having teeth on one of its lateral faces; (d) a clutch pinion rotatably engaged with said shaft and mounted to move in translation on said shaft, between a disengaged position in which said first pinion and said clutch pinion are at a distance and an engaged position in which said pinion and said clutch pinion are in contact, said clutch pinion having teeth on its face directed toward said lateral face of said first pinion provided with teeth; (e) whereby said first pinion and said clutch pinion are connected for transmitting the rotary movement of said shaft to said first pinion, the translatory movement of said clutch pinion in the axial direction of said shaft being controlled by a push-rod connected to a vertically moving actuating member which is maneuverable from the upper part of the apparatus. 2. Rack and pinion mechanism according to claim 1, wherein said actuating member of said push-rod comprises a sleeve coaxial with said control shaft connected at its upper part to a magnetic means and retained in a raised position, in which said push-rod keeps said clutch pinion in an engaged position, by a magnetic coil, stopping of the supply of this magnetic coil causing said sleeve to descend and said push-rod and said clutch pinion to move, so as to achieve emergency shutdown and descent of said control rod under the effect of its own weight. 3. Rack and pinion mechanism according to claim 2, comprising a return spring associated with said sleeve so as to return it to a lower position, said push-rod returning said clutch pinion back into disengaged position. 4. Rack and pinion mechanism according to claim 1, wherein said drive means comprises a three-phase motor locked in position by braking by lack of current disposed outside a sealed casing in communication with the reactor vessel and containing said rack and pinion mechanism, said motor being associated, for transmitting rotary movement to said rack and pinion mechanism, with a permanent magnetic coupling.
	description	FIG. 1 is described first. The invention is intended for operating a grab 1 in charge of picking up samples of waste tipped beforehand into a tank 2 of a silo 3 whose shape is roughly that of a bottle, with a narrow opening 4 at the top. A concrete slab 5 covers the silo 3 and protects the outside from the irradiation which might be caused by the waste; however it can be freed at the place of the opening 4. The grab 1 is suspended from a cable 6 which reaches as high as an overhead crane 7 astride the slab 5 and the opening 4 and a pulley 8 of this crane 7, where it is turned laterally towards a second pulley 9, then towards a motorised winch 10 not shown in detail. This equipment makes it possible to lower the grab 1 down to the surface of the waste but is not able to displace it laterally in the tank 2; the articulated arm described below fills this gap. It carries the general reference 11 and essentially comprises an upper section 12 lowered through the opening 4 and a lower section 13 linked to the preceding section by an articulation 14 and which extends lower into the tank 2; its free end carries a pulley 15 over which passes cable 6. The grab 1 thus hangs vertically from this pulley 15 and can be set at any position in the tank 2 depending on the angle imposed between sections 12 and 13 by controlling the articulation 14. It is to be noted that the latter is equipped with another pulley 16, under which cable 6 passes so as to maintain it fairly closely in the alignment of each of the sections 12 and 13 and to avoid it rubbing against the edge of the opening 4. Referring to FIG. 2, it can be seen that the lower section 13 consists of two parallel girders 17 whose spacing is sufficient to allow the grab 1 to be lowered between them (its external outline is marked). The articulation 14 is oriented in such a way that the lower section pivots in the vertical and radial planes (passing along the central axis of the opening 4). Referring now to FIG. 3, it can be seen that slab 5 carries a frame 18 around the opening 4, which comprises in particular a fixed ring 19 with ball-bearings 20, whose other ring 21, turning, is integral with a crown 22 on which the support structure 23 can be set and fixed by a circle of lateral screws 24 and vertical screws 25. The arm 11 is suspended from the support structure 23. A pinion gear 26 of a drive motor 27 fixed on the support structure 23 then engages with a toothed crown 28 cut on the internal surface of the fixed ring 19: it results from this construction that starting up the motor 27 displaces the pinion 26 and consequently the support structure 23 and the articulated arm 11 as a whole along the toothed wheel 28 and around the opening 4, which places the lower section 13 and the grab 1 in the angular direction required. The support structure 23 comprises another motor 29, which moves a series of gears 30 and displaces an endless chain 31 at its end; the chain 31 extends along the upper section 12 as far as a pinion 32 set at the articulation 14 and integral with the lower section 13: this is the mechanism which varies the angle of articulation 14 and displaces the lower section 13 and the pulley 15 towards the walls of the silo 2 by the amount required. FIG. 2 shows that the chain 31 and the pinion 32 can be double and located on either side of the articulation pulley 16, which balances the efforts of the girders 17. The chain 31 presents the advantages over other drives of resisting mechanical strains and irradiation better, and the possibility of being assembled and disassembled quickly. In FIG. 1 a plug 33 in two parts is to be noted, the smallest of which 34 has the cross-section of the upper section 12; the plug 33 is installed over the opening 4 to close it, completely in the absence of the articulated arm 11, but with the part 34 taken off when the articulated arm 11 is placed through the opening 4 and left immobile; it then adjusts itself around the upper section 12 and maintains an almost perfect seal. Another precaution for maintaining the seal around the opening 4 consists of adding a metallic skirt 35, circular and continuous under the fixed ring 19 with the base dipped into liquid poured into a circular metallic groove 36 with cross-section in the shape of a trough and which is fixed on the slab 5 around the opening 4 (FIG. 3). The assembly begins by installing the frame 18 and the crown 22 around the opening 4, then the articulated arm 11 is lowered through the opening 4 and its support structure 23 is screwed to the crown 22. The lower section 13 is pivoted to the back, in an almost horizontal position, directed towards the left, as represented by dots and dashes in FIG. 1, before lowering the grab 1 and passing it between the girders 17. An opposite pivoting of the lower section 13, towards the front, makes it possible for the suspension pulley 15 to intercept and catch the cable 6. The grab 1 is finally brought above the desired point in the tank to take a sample, by displacing the articulated arm 11 around the opening 4 to place the lower section 13 at the orientation desired and making it pivot to place the grab 1 at the distance desired from the opening 4 to thus provide guidance in polar coordinates. The ability to force the pivoting backwards, for the passage of the grab 1, and forwards for guiding it, is an important aspect of the device, due to the control reversibility of the movement by the chain 31. It is indicated that the lower section 13 is to be constructed in two parts articulated together, which will now be described referring to FIG. 4. These two parts carry the references 41 and 42 and are linked together by an articulation axle 43 placed between their front edges 44. During the descent and the assembly of the arm 11 they are folded against each other and against the upper section 12, the upper part 41 then being directed upwards and the whole of the articulated arm 11 forming roughly an xe2x80x9cNxe2x80x9d with moderate height. When the articulated arm 11 has been assembled, the lower section 13 is lowered into the tank 2 while pivoting; the upper part 41, fairly short, passes freely through the opening 4. The parts 41 and 42 end up by being extended and touch each other, after which they can be locked. A locking system proposed here comprises at least one pin 46 set at the bottom of the upper part 41, near to the rear edge 47, opposite the articulation 43. The pin 46 is directed downwards, and pushed in this direction by a spring, not shown, which makes it protrude outside the upper part 41. But the action of the spring is countered by a cable 51 hooked at the back of the pin 46 and which extends first of all around the articulation pulley 16, then along the upper section 12 before ending above the slab 5. The cable 51 can be caught by a telemanipulator or otherwise, and pulled to overcome the action of the spring 48 and return the pin 46 into the upper part 41 of the lower section 13. This is what is done when the parts 41 and 42 are extended at assembly, so that the pin 46 does not interfere with the approach of the rear edges 47 of the two parts 41 and 42; then the cable 51 is released so that the pin 46 protrudes from the upper section 41 and enters a notch 52 made in the lower part 42 and which then opens in front of it. The pin 46 now becomes integral with the parts 41 and 42 and counteracts any rotation around the articulation 43 until the cable 51 is pulled again. The locking and unlocking action can be made automatic so as not to rely on a cable 51 ending outside the opening 4 and thus susceptible to compromising the seal. The cable used then corresponds only to the rear portion 53 of the cable 51, which is hooked at its end opposite to the pin 46 to an articulation point 14 linked to the upper section 12 in such a way that, when the upper section 13 pivots to its maximum, the cable portion 53 winds itself up around the articulation 14, stretches itself and pulls the pin 46 in the same way as before; the cable portion 53 extends when the lower section 13 is sent slightly lower, which releases the pin 46 and makes it possible for it to enter the notch 52 and to ensure the linkage between the parts 41 and 42 as before. The tension of the portion of cable 53 is only produced in the extreme pivoting positions towards the front, in the angle xcex2 in FIG. 4. The angle xcex1 corresponds to the pivoting range in which the cable 53 is released and or the locking of the parts 41 and 42 can be maintained. A cam, not shown but of the generally known type, and which is put into movement by the tension of the cable portion 53, makes it possible alternatively to let pin 46 protrude or to maintain it within the part 41 once the cable portion 53 has been released, and thus either to lock the parts 41 and 42, or not. Thus it suffices to bring the lower section 13 to its extreme front position in the angle xcex2 after assembling the articulated arm 11 to lock the parts 41 and 42, then to bring it back once to control the unlocking at the end of operations, and the parts 41 and 42 fold together as soon as the lower section 13 is brought backwards past the vertical. The articulated arm 11 can then be dismantled from the opening 4. The grab 1 has been lifted and taken away before unlocking and after moving the lower section 13 backwards, in extension position for the parts 41 and 42, to free the cable 6.
	summary
054105773	summary	TECHNICAL FIELD This invention relates to the field of nuclear reactors and core meltdown damage prevention and more specifically to core catchers. BACKGROUND ART In the field of water cooled nuclear reactors it is well known that the most probable accident in the nuclear reactor is a reactor core meltdown which occurs when the capability to remove heat from the nuclear reactor core is lost. When a reactor core meltdown occurs, the core melts through the pressure vessel and the molten material drops onto the containment floor, which is typically fabricated from a concrete material. The reactor is surrounded by a containment building which is intended to prevent the radioactivity from escaping into the environment. The resultant mixture of molten and solid core material from a water cooled nuclear reactor core melt accident is "corium". Corium results from a zirconium clad fuel and is generally comprised of uranium oxide (UO.sub.2), steel (Fe), zirconium oxide (ZrO.sub.2), zirconium (Zr) and fission products. In a typical coremelt accident, the molten corium core is an uncoolable, high power density mass which exits the bottom of the reactor pressure vessel. The geometry of the molten corium is ill defined. The high temperature corium can penetrate the containment floor by high temperature decomposition of the concrete. When this occurs, the reaction between corium and the concrete generates flammable, noncondensable gases. More specifically, the decomposition of concrete generates carbon dioxide (CO.sub.2). The resultant carbon dioxide reacts with the zirconium in corium to generate carbon monoxide (CO). Further, zirconium reacts with water in the containment and in the concrete to generate hydrogen (H.sub.2). The noncondensable gases, hydrogen and carbon monoxide, may overpressurize the reactor containment causing its failure. Also, the hydrogen and carbon monoxide may burn or detonate with air in the containment building which could cause containment failure. Furthermore, with failure of the containment, the potential for leakage of radioactive material to the environment is extremely high. Many methods for preventing containment failure have been proposed. Generally, these methods require spreading the molten material out so it can cool and freeze by flooding the underreactor cavity with water. The basic drawback to all such approaches is ensuring the distribution or geometry of the molten core material during a core-melt accident. For example, lead core catchers have been developed in an effort to solve the problems involved with a core-melt accident. Lead has a sufficiently high density to float the core debris but there is no way to absorb the core debris such that the power density of the molten core is lowered. Further, a lead core catcher can not prevent reaction of zirconium in corium with the water which drains down onto the corium with time. A core-melt source reduction system has been developed for a gas cooled fast reactor which utilizes a stainless steel clad fuel. The major difference between the system developed for the stainless steel clad fuel and the present invention for a zirconium clad fuel is that stainless steel does not react with most materials in a core-melt accident. Unlike stainless steel, zirconium is highly reactive. Because of this highly reactive nature, there will be fundamental compositional differences between the system developed for the stainless steel clad fuel and that for zirconium clad fuel. Therefore, it is an object of this invention to provide a core-melt source reduction system which stops the progression of a high temperature core through the containment floor during a core-melt accident. It is another object of the present invention to provide a core-melt source reduction system which prevents the generation of noncondensable gases when the core materials react with the containment floor. It is a further object of the present invention to provide a core-melt source reduction system which does not require any assumptions about the geometry or timing of the molten core material in a core-melt accident. It is yet another object of the present invention to provide such a core-melt source reduction system which can be incorporated into or replace the existing containment floor. Further, it is object of the present invention to provide a core-melt source reduction system which minimizes heat rejection to the containment floor early in the core-melt accident. It is yet another object of the present invention to provide a core-melt source reduction system which ends the accident sequence with a long term, cold, stable state. It is a further object of the present invention to provide a core-melt source reduction system which traps radionuclides in a solidified matrix. DISCLOSURE OF THE INVENTION Other objects and advantages will be accomplished by the present invention which serves to provide a system which can contain a molten core after a nuclear reactor core meltdown accident such that containment failure is prevented. The core-melt source reduction system, hereinafter "the system", of the present invention includes alternate layers of two different materials as part of the floor under the reactor. The first material reacts with and absorbs the core material in a manner such that the core spreads out and the progression of the core is slowed and ultimately halted. The second material acts as a barrier to slow the descent of the molten core through a layer of the first material to provide time for dissolution of the core materials within the first material. During the reaction between the molten core and the layers of the first and second materials, no noncondensable gases are generated such that the containment will not fail as a result of overpressurization. Further, the resultant material is a solidified thermally hot stable waste material which can ultimately be cooled to a stable state matrix. The stable state matrix is such that radionuclides are trapped within the matrix to minimize the consequences of leakage.

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